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Sample records for reactor 1 experiment

  1. Reactor G1: high power experiments

    International Nuclear Information System (INIS)

    Laage, F. de; Teste du Baillet, A.; Veyssiere, A.; Wanner, G.

    1957-01-01

    The experiments carried out in the starting-up programme of the reactor G1 comprised a series of tests at high power, which allowed the following points to be studied: 1- Effect of poisoning by Xenon (absolute value, evolution). 2- Temperature coefficients of the uranium and graphite for a temperature distribution corresponding to heating by fission. 3- Effect of the pressure (due to the coiling system) on the reactivity. 4- Calibration of the security rods as a function of their position in the pile (1). 5- Temperature distribution of the graphite, the sheathing, the uranium and the air leaving the canals, in a pile running normally at high power. 6- Neutron flux distribution in a pile running normally at high power. 7- Determination of the power by nuclear and thermodynamic methods. These experiments have been carried out under two very different pile conditions. From the 1. to the 15. of August 1956, a series of power increases, followed by periods of stabilisation, were induced in a pile containing uranium only, in 457 canals, amounting to about 34 tons of fuel. A knowledge of the efficiency of the control rods in such a pile has made it possible to measure with good accuracy the principal effects at high temperatures, that is, to deal with points 1, 2, 3, 5. Flux charts giving information on the variations of the material Laplacian and extrapolation lengths in the reflector have been drawn up. Finally the thermodynamic power has been measured under good conditions, in spite of some installation difficulties. On September 16, the pile had its final charge of 100 tons. All the canals were loaded, 1,234 with uranium and 53 (i.e. exactly 4 per cent of the total number) with thorium uniformly distributed in a square lattice of 100 cm side. Since technical difficulties prevented the calibration of the control rods, the measurements were limited to the determination of the thermodynamic power and the temperature distributions (points 5 and 7). This report will

  2. Reactor G1: high power experiments; Experiences a forte puissance

    Energy Technology Data Exchange (ETDEWEB)

    Laage, F de; Teste du Baillet, A; Veyssiere, A; Wanner, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Retel, H [Societe Rateau, D.E.A. (France)

    1957-07-01

    The experiments carried out in the starting-up programme of the reactor G1 comprised a series of tests at high power, which allowed the following points to be studied: 1- Effect of poisoning by Xenon (absolute value, evolution). 2- Temperature coefficients of the uranium and graphite for a temperature distribution corresponding to heating by fission. 3- Effect of the pressure (due to the coiling system) on the reactivity. 4- Calibration of the security rods as a function of their position in the pile (1). 5- Temperature distribution of the graphite, the sheathing, the uranium and the air leaving the canals, in a pile running normally at high power. 6- Neutron flux distribution in a pile running normally at high power. 7- Determination of the power by nuclear and thermodynamic methods. These experiments have been carried out under two very different pile conditions. From the 1. to the 15. of August 1956, a series of power increases, followed by periods of stabilisation, were induced in a pile containing uranium only, in 457 canals, amounting to about 34 tons of fuel. A knowledge of the efficiency of the control rods in such a pile has made it possible to measure with good accuracy the principal effects at high temperatures, that is, to deal with points 1, 2, 3, 5. Flux charts giving information on the variations of the material Laplacian and extrapolation lengths in the reflector have been drawn up. Finally the thermodynamic power has been measured under good conditions, in spite of some installation difficulties. On September 16, the pile had its final charge of 100 tons. All the canals were loaded, 1,234 with uranium and 53 (i.e. exactly 4 per cent of the total number) with thorium uniformly distributed in a square lattice of 100 cm side. Since technical difficulties prevented the calibration of the control rods, the measurements were limited to the determination of the thermodynamic power and the temperature distributions (points 5 and 7). This report will

  3. Development of a reactor thermalhydraulic experiment databank(SORTED1)

    International Nuclear Information System (INIS)

    Bang, Young Seck; Kim, Eun Kyoung; Kim, Hho Jung; Lee, Sang Yong

    1994-01-01

    The recent trend in thermalhydraulic safety analysis of nuclear power plant shows the best-estimate and probabilistic approaches, therefore, the verification of the best-estimate code based on the applicable experiment data has been required. The present study focused on developing a simple databank, SORTED1, to be effectively used for code verification. The development of SORTED1 includes a data collection from the various sources including ENCOUNTER, which is the reactor safety data bank of U.S. Nuclear Regulatory Commission, a reorganization of collected resources suitable for requirements of SORTED1 database management system (DBMS), and a development of a simple DBMS. The SORTED1 is designed in Unix environment with graphic user interface to improve a user convenience and has a capability to provide the test related information. The currently registered data in SORTED1 cover 759 thermalhydraulic tests including LOFT, Semiscale, etc

  4. Experiment on continuous operation of the Brazilian IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Freitas Pintaud, M. de

    1994-01-01

    In order to increase the radioisotope production in the IEA-R1 research reactor at IPEN/CNEN-SP, it has been proposed a change in its operation regime from 8 hours per day and 5 days per week to continuous 48 hours per week. The necessary reactor parameters for this new operation regime were obtained through an experiment in which the reactor was for the first time operated in the new regime. This work presents the principal results from this experiment: xenon reactivity, new shutdown margins, and reactivity loss due to fuel burnup in the new operation regime. (author)

  5. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

    Edwards, R.M.; Power, M.A.; Bryan, M.

    1992-01-01

    A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

  6. A data base for reactor physics experiments at KUCA, 1

    International Nuclear Information System (INIS)

    Ichihara, Chihiro; Hayashi, Masatoshi; Fujine, Shigenori; Wakamatsu, Susumu.

    1986-01-01

    A data base of the experiment done at the Critical Assembly of Kyoto University(KUCA) was constructed both on personal computers and a main frame. A retrieval data base based on each experiment serve as the key data base. The critical experiment data, geometries of the core configuration or fuel elements, and the various numeric data are referred after the results of the retrieval. The personal computer program for this data base is made using BASIC language and the whole system consist of the retrieval data base and the graphic data. The construction of the critical experiment data is now in progress. The data base system can be supplied to the KUCA users with floppy disks. A universal information retrieval system, FAIRS is prepared at the Data Processing Center Kyoto University. By using this system, the retrieval data base of the experiment was constructed. The image information such as core configuration and fuel elements are stored by using ELF system which can be linked to the FAIRS. The data base on FAIRS can be referred from each university through an online network. However, ELF is a closed service within Kyoto University at present. (author)

  7. The Dragon reactor experiment

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    The concept on which the Dragon Reactor Experiment was based was evolved at the Atomic Energy Research Establishment at Harwell in 1956, and in February of that year a High Temperature Gas- cooled Reactor Project Group was set up to study the feasibility of a helium-cooled reactor with a graphite or beryllium moderator, and with the emphasis on the thorium fuel cycle [af

  8. Reactor dynamics experiment of N.S. Mutsu using pseudo random signal. 1

    International Nuclear Information System (INIS)

    Hayashi, Koji; Nabeshima, Kunihiko; Shinohara, Yoshikuni; Shimazaki, Junya; Inoue, Kimihiko; Ochiai, Masaaki.

    1993-10-01

    In order to investigate dynamics of the reactor plant of the nuclear ship Mutsu, reactor noise experiments using pseudo random binary sequences (PRBS) have been planned, and a preliminary experiment was performed on March 4, 1991 in the first experimental navigation with the aim of checking the experimental procedures and conditions. The experiments using both reactivity and load disturbances were performed at 70 % of reactor power and under a quiet sea condition. Each PRBS was applied by manual operation of the control rod or the main steam valve. Various signals of the plant responses and of the acceleration of ship motion were measured. From the results obtained, we confirmed that (1) the procedures and experimental conditions determined prior to the experiment were suitable for performing the PRBS experiments, (2) when the PRBS disturbances were applied, the plant state remained quite stable, and (3) the quality of the measured data is adequate for the purpose of dynamics analysis. This paper summarizes the planning and preparation of the experiment, the instruction for the experiment and logs, the data recording conditions, recorded signal wave forms and the results of power spectral analysis. (author)

  9. Mark I 1/5-scale boiling water reactor pressure suppresion experiment quick-look report

    International Nuclear Information System (INIS)

    Lai, W.; Collins, E.K.

    1977-01-01

    This report is intended as a ''quick-look'' report summarizing the experimental results obtained from pressure suppression experiment numbers 2.1, 2.2, and 2.3 that were performed on the Lawrence Livermore Laboratory's 1/5-scale boiling water reactor (BWR) Mark I pressure suppression experimental facility on April 26, 1977. A brief description of the general nature of the tests and a summary of the actual tests that were performed are given

  10. Fission product monitoring of TRISO coated fuel for the advanced gas reactor-1 experiment

    International Nuclear Information System (INIS)

    Scates, Dawn M.; Hartwell, John K.; Walter, John B.; Drigert, Mark W.; Harp, Jason M.

    2010-01-01

    The US Department of Energy has embarked on a series of tests of TRISO coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burnup of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B's) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  11. The reactor kinetics code tank: a validation against selected SPERT-1b experiments

    International Nuclear Information System (INIS)

    Ellis, R.J.

    1990-01-01

    The two-dimensional space-time analysis code TANK is being developed for the simulation of transient behaviour in the MAPLE class of research reactors. MAPLE research reactor cores are compact, light-water-cooled and -moderated, with a high degree of forced subcooling. The SPERT-1B(24/32) reactor core had many similarities to MAPLE-X10, and the results of the SPERT transient experiments are well documented. As a validation of TANK, a series of simulations of certain SPERT reactor transients was undertaken. Special features were added to the TANK code to model reactors with plate-type fuel and to allow for the simulation of rapid void production. The results of a series of super-prompt-critical reactivity step-insertion transient simulations are presented. The selected SPERT transients were all initiated from low power, at ambient temperatures, and with negligible coolant flow. Th results of the TANK simulations are in good agreement with the trends in the experimental SPERT data

  12. Acoustic detection of boiling in the Sodium Loop Safety Facility in-reactor experiment P1

    International Nuclear Information System (INIS)

    Carey, W.M.; Anderson, T.T.; Bobis, J.P.

    1976-06-01

    Acoustic data were obtained from two high-temperature lithium niobate microphones on the loop background noise and transient pressure pulses during the Sodium Loop Safety Facility (SLSF) P1 in-reactor experiment. This experiment simulated an LMFBR loss-of-piping-integrity (LOPI) transient on a nineteen element, end-of-life, enriched-UO 2 fuel assembly. The microphones were exposed to liquid sodium at a distance 4.85 meters above the reactor core at temperatures between 315 0 and 590 0 C. The distance and location of the microphones in the P1 Test Train provided an attenuative transmission path which was undesirable for optimum acoustic detection of sodium boiling and fuel failure. The data gathered on the loop background noise was observed to be dominated by pump and electrical noise at frequencies below 1.5 KHz and appeared to be dominated by flow induced local turbulence noise at higher frequencies. During the period of time that the sodium in the fuel assembly was at its saturation temperature 943 0 C (1730 0 F), as indicated by the wire wrap thermocouples, several discrete pulses were observed with peak-to-peak pressure between 3.3 kPa and 7.9 kPa and center frequencies between 360 and 550 Hz. The pulses occurred at two separate gradually increasing repetition rates. These observations appear to be consistent with the result of an impulsive forcing function interacting with a band passed Helmholtz resonator. These data are consistent with the hypothesis that sodium boiling occurred in the P1 fuel assembly, resulting in the formation of individual voids that collapsed upon reaching the subcooled sodium. These data provide pertinent information regarding the feasibility of sodium boiling detection and may provide additional insight into the dynamics of the void behavior

  13. Irradiation experience of IPEN fuel at IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Perrotta, Jose A.; Neto, Adolfo; Durazzo, Michelangelo; Souza, Jose A.B. de; Frajndlich, Roberto

    1998-01-01

    IPEN/CNEN-SP produces, for its IEA-R1 Research Reactor, MTR fuel assemblies based on U 3 O 8 -Al dispersion fuel type. Since 1985 a qualification program on these fuel assemblies has been performed. Average 235 U burnup of 30% and peak burnup of 50% was already achieved by these fuel assemblies. This paper presents some results acquire, by these fuel assemblies, under irradiation at IEA-R1 Research Reactor. (author)

  14. CANDU reactor experience: fuel performance

    International Nuclear Information System (INIS)

    Truant, P.T.; Hastings, I.J.

    1985-07-01

    Ontario Hydro has more than 126 reactor-years experience in operating CANDU reactors. Fuel performance has been excellent with 47 000 channel fuelling operations successfully completed and 99.9 percent of the more than 380 000 bundles irradiated operating as designed. Fuel performance limits and fuel defects have had a negligible effect on station safety, reliability, the environment and cost. The actual incapability charged to fuel is less than 0.1 percent over the stations' lifetimes, and more recently has been zero

  15. IEA-R1 Nuclear Research Reactor: 58 Years of Operating Experience and Utilization for Research, Teaching and Radioisotopes Production

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Jose Patricio Nahuel; Filho, Tufic Madi; Saxena, Rajendra; Filho, Walter Ricci [Nuclear and Energy Research Institute, IPEN-CNEN/SP, Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP, Av. Prof. Lineu Prestes 2242 Cid Universitaria CEP: 05508-000- Sao Paulo-SP (Brazil)

    2015-07-01

    IEA-R1 research reactor at the Instituto de Pesquisas Energeticas e Nucleares (Nuclear and Energy Research Institute) IPEN, Sao Paulo, Brazil is the largest power research reactor in Brazil, with a maximum power rating of 5 MWth. It is being used for basic and applied research in the nuclear and neutron related sciences, for the production of radioisotopes for medical and industrial applications, and for providing services of neutron activation analysis, real time neutron radiography, and neutron transmutation doping of silicon. IEA-R1 is a swimming pool reactor, with light water as the coolant and moderator, and graphite and beryllium as reflectors. The reactor was commissioned on September 16, 1957 and achieved its first criticality. It is currently operating at 4.5 MWth with a 60-hour cycle per week. In the early sixties, IPEN produced {sup 131}I, {sup 32}P, {sup 198}Au, {sup 24}Na, {sup 35}S, {sup 51}Cr and labeled compounds for medical use. During the past several years, a concerted effort has been made in order to upgrade the reactor power to 5 MWth through refurbishment and modernization programs. One of the reasons for this decision was to produce {sup 99}Mo at IPEN. The reactor cycle will be gradually increased to 120 hours per week continuous operation. It is anticipated that these programs will assure the safe and sustainable operation of the IEA-R1 reactor for several more years, to produce important primary radioisotopes {sup 99}Mo, {sup 125}I, {sup 131}I, {sup 153}Sm and {sup 192}Ir. Currently, all aspects of dealing with fuel element fabrication, fuel transportation, isotope processing, and spent fuel storage are handled by IPEN at the site. The reactor modernization program is slated for completion by 2015. This paper describes 58 years of operating experience and utilization of the IEA-R1 research reactor for research, teaching and radioisotopes production. (authors)

  16. Air scaling and modeling studies for the 1/5-scale mark I boiling water reactor pressure suppression experiment

    Energy Technology Data Exchange (ETDEWEB)

    Lai, W.; McCauley, E.W.

    1978-01-04

    Results of table-top model experiments performed to investigate pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peach Bottom Mark I boiling water reactor containment system guided subsequent conduct of the 1/5-scale torus experiment and provided new insight into the vertical load function (VLF). Pool dynamics results were qualitatively correct. Experiments with a 1/64-scale fully modeled drywell and torus showed that a 90/sup 0/ torus sector was adequate to reveal three-dimensional effects; the 1/5-scale torus experiment confirmed this.

  17. Air scaling and modeling studies for the 1/5-scale mark I boiling water reactor pressure suppression experiment

    International Nuclear Information System (INIS)

    Lai, W.; McCauley, E.W.

    1978-01-01

    Results of table-top model experiments performed to investigate pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peach Bottom Mark I boiling water reactor containment system guided subsequent conduct of the 1/5-scale torus experiment and provided new insight into the vertical load function (VLF). Pool dynamics results were qualitatively correct. Experiments with a 1/64-scale fully modeled drywell and torus showed that a 90 0 torus sector was adequate to reveal three-dimensional effects; the 1/5-scale torus experiment confirmed this

  18. Experience and research with the IEA-R1 Brazilian reactor

    International Nuclear Information System (INIS)

    Fulfaro, R.; Sousa, J.A. de; Nastasi, M.J.C.; Vinhas, L.A.; Lima, F.W.

    1982-06-01

    The IEA-R1 reactor of the Instituto de Pesquisas Energeticas e Nucleares, IPEN, of Sao Paulo, Brazil, a lightwater moderated swimming-pool research reactor of MTR type, went critical for the first time on September 16, 1957. In a general way, in these twenty four years the reactor was utilized without interruption by users of IPEN and other institutions, for the accomplishment of work in the field of applied and basic research, for master and doctoral thesis and for technical development. Some works performed and the renewal programme established for the IEA-R1 research reactor in which several improvements and changes were made. Recent activities in terms of production of radioisotopes and some current research programm in the field of Radiochemistry are described, mainly studies and research on chemical reactions and processes using radioactive tracers and development of radioanalytical methods, such as neutron activation and isotopic dilution. The research programmes of the Nuclear Physics Division of IPEN, which includes: nuclear spectroscopy studies and electromagnetic hyperfine interactions; neutron diffraction; neutron inelastic scattering studies in condensed matter; development and application of the technique of fission track register in solid state detectors; neutron radioactive capture with prompt gamma detection and, finally, research in the field of nuclear metrology, are presented. (Author) [pt

  19. Experience and research with the IEA-R1 Brazilian reactor

    International Nuclear Information System (INIS)

    Fulfaro, R.; Sousa, J.A. de; Nastasi, M.J.C.; Vinhas, L.A.; Lima, F.W. de.

    1982-06-01

    The IEA-R1 reactor of the Instituto de Pesquisas Energeticas e Nucleares, IPEN, of Sao Paulo, Brazil, a lighwater moderated swimming-pool research reactor of MTR type, went critical for the first time on September 16, 1957. In a general way, in these twenty four years the reactor was utilized without interruption by users of IPEN and other institutions, for the accomplishment of work in the field of applied and basic research, for master and doctoral thesis and for technical development. Some works performed and the renewal programme established for the IEA-R1 research reactor in which several improvements and changes were made. Recent activities in terms of production of radioisotopes and some current research programm in the field of Radiochemistry are described, mainly studies and research on chemical reactions and processes using radioactive tracers and development of radioanalytical methods, such as neutron activation and isotopic dilution. It is also presented the research programmes of the Nuclear Physics Division of IPEN, which includes: nuclear spectroscopy studies and electromagnetic hyperfine interactions; neutron diffraction; neutron inelastic scattering studies in condensed matter; development and application of the technique of fission track register in solid state detectors; neutron radioactive capture with prompt gamma detection and, finally, research in the field of nuclear metrology. (Author) [pt

  20. Physical experiments. Reactor theory

    International Nuclear Information System (INIS)

    Korn, H.; Werle, H.; Bluhm, H.; Fieg, G.; Kappler, F.; Kuhn, D.; Lalovic, M.; Woll, D.; Kuefner, K.; Woznicki, Z.; Buckel, G.; Stehle, B.; Borgwaldt, H.

    1975-01-01

    The γ-spectrum in SNEAK 9C-1 and 9C-2 was measured by means of Si(Li) solid state detectors for verification of methods of shielding calculation. The blanket spectra turned out to be slightly harder than the spectra in the fissile zone; the plutonium spectra are slightly harder than the respective uranium spectra. This result is expected to be explained by studies to be carried out on the basis of a γ-transport program. For reactor theoretical calculations two 2-dimensional diffusion programs were compared with each other, and a 3-dimensional diffusion program was compared with a flux synthesis program. An improved source iteration scheme was drafted for the Karlsruhe Monte Carlo code. (orig.) [de

  1. Experience gained in refurbishing of the ET-R R-1 reactor in Egypt

    Energy Technology Data Exchange (ETDEWEB)

    Khattab, M; Dimitri, F; Chaath, K [Reactor department, nuclear research center atomic energy authority, Cairo, (Egypt)

    1995-10-01

    This paper describes the in-service program and rehabilitation plan of the control, measuring instrumentation and radiation monitoring equipment as well as the computerized safety logic and signaling systems. the in-service program includes reactor core and pressure vessels. Spent fuel tank and primary cooling circuit have been inspected. Current problems and future plan for improving the safety systems are discussed. 10 figs., 1 tab.

  2. Experience gained in refurbishing of the ET-R R-1 reactor in Egypt

    International Nuclear Information System (INIS)

    Khattab, M.; Dimitri, F.; Chaath, K.

    1995-01-01

    This paper describes the in-service program and rehabilitation plan of the control, measuring instrumentation and radiation monitoring equipment as well as the computerized safety logic and signaling systems. the in-service program includes reactor core and pressure vessels. Spent fuel tank and primary cooling circuit have been inspected. Current problems and future plan for improving the safety systems are discussed. 10 figs., 1 tab

  3. Security devices and experiment facilities at ENEA TRIGA RC-1 reactor

    International Nuclear Information System (INIS)

    Bianchi, P.; Festinesi, A.; Santoro, E.; Tardani, G.; Magli, M.; Reis, G.

    1990-01-01

    RC-1 TRIGA operating exercise staff has produced some auxiliary security devices. These are the neutron source automatic handling device, irradiated samples rabbit connection rotating rack, and auxiliary equipment for transferring hot fuel elements. The reactor electronic control instrumentation system includes various instrumentation channels, the operating capability of which must be verified by the licensee as per Italian regulations. In order to obtain automatic and repeatable operations, TEMAV designed and constructed a remotely-driven source transfer device, based on requirements, performance specifications and technical data supplied by ENEA-TIB. The pneumatic irradiating system for short lived materials allows extraction of radiated samples in a time no longer than 4 seconds. To optimize the system, both as to operability and health protection, a specific rotating rack for the connection of irradiated samples with pneumatic transfer (RABBIT) was produced. To permit 1 MW hot fuel element storage in pits it is necessary to remove hot 100 KW fuel elements and transfer them to a re-treatment plant. Feasibility studies showed the impossibility of using heavy trucks inside the reactor hall. To avoid problems trucks are left outside the reactor hall and only the PEGASO container is removed with a special device that runs on rails. Movement from Rail truck is assured by an electromotor driving pull device and security cable

  4. Natural uranium-graphite system. Critial experiments on the G1 reactor; Systeme uranium naturel-graphite. Experiences critiques sur le reacteur G1

    Energy Technology Data Exchange (ETDEWEB)

    Schmitt, A P; Tanguy, P; Teste du Bailler, A; Zaleski, C P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    A number of experiments have been performed during the start up period of the G1 (1956) and G2 (1958) reactors in Marcoule, both on their lattices and on different lattices (hollow rods, clusters, under moderated lattices). The first chapter gives a thorough description of the two reactors. The second chapter deals with buckling measurements, both absolute (flux plots) and relative by the method of progressive substitution. The experimental results are summarised in Table VI. The third chapter contains a number of other measurements performed on G1. (author)Fren. [French] Le demarrage des reacteurs G1 (1956) et G2 (1958) de Marcoule nous a permis d'effectuer une serie d'experiences tant sur les reseaux de ces piles que sur des reseaux differents (elements tubulaires ou divises, reseaux sous-moderes, etc...). Dans une premiere partie, nous donnons une description detaillee des deux reacteurs. Dans la deuxieme partie, relative aux mesures de laplaciens, nous decrivons d'abord les mesures absolues de laplaciens (cartes de flux), puis les mesures relatives effectuees par la methode originale de remplacement progressif. Les resultats experimentaux sont rassembles dans le tableau VI. Dans la troisieme partie, nous rappelons un certain nombre d'autres mesures effectuees sur G1. (auteur)

  5. Qualification of JEFF3.1.1 library for high conversion reactor calculations using the ERASME/R experiment

    Energy Technology Data Exchange (ETDEWEB)

    Vidal, J. F.; Noguere, G.; Peneliau, Y.; Santamarina, A. [CEA, DEN, DER/SPRC/LEPh, Cadarache, F-13108 Saint-Paul-lez-Durance (France)

    2012-07-01

    With its low CO{sub 2} production, Nuclear Energy appears to be an efficient solution to the global warming due to green-house effect. However, current LWR reactors are poor uranium users and, pending the development of Fast Neutron Reactors, alternative concepts of PWR with higher conversion ratio (HCPWR) are being studied again at CEA, first studies dating from the middle 80's. In these French designs, low moderation ratio has been performed by tightening the lattice pitch, achieving a conversion ratio of 0.8-0.9 with a MOX fuel coming from PWR UOX recycling. Theses HCPWRs are characterized by a harder neutron spectrum and the calculation uncertainties on the fundamental neutronics parameters are increased by a factor 3 regarding a standard PWR lattice, due to the major contribution of the Plutonium isotopes and of the epithermal energy range to the reaction rates. In order to reduce these uncertainties, a 3-year experimental validation program called ERASME has been performed by CEA from 1984 to 1986 in the EOLE reactor. Monte Carlo analysis of the ERASME/R experiments with the Monte Carlo code TRIPOLI4 allowed the qualification of the recommended JEFF.3.1.1 library for major neutronics parameters. K{sub eff} of the MOX under-moderated lattice is over-predicted by 440 {+-} 830 pcm (2{sigma}); the conversion ratio, indicator of the good use of uranium, is also slightly over-predicted: 2 % {+-} 4 % (2{sigma}) and the same for B4C absorber rods worth and soluble boron worth, over-predicted by 2 %, both in the 2 standard deviations range. The radial fission maps of heterogeneities (water-holes, B4C and fertile rods) are well reproduced: maximal (C-E)/E dispersion is 1.3 %, maximal power peak error is 2.7 %. The void reactivity worth is the only parameter poorly calculated with an overprediction of +12.4% {+-} 1.5%. ERASME/R analysis of MOX reactivity, void effect and spectral indexes will contribute to the reevaluation of {sup 241}Am and Plutonium isotopes

  6. Experience on wet storage spent fuel sipping at IEA-R1 Brazilian research reactor

    International Nuclear Information System (INIS)

    Perrotta, J.A.; Terremoto, L.A.A.; Zeituni, C.A.

    1998-01-01

    The IEA-R1 research reactor of the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) is a pool type reactor of B and W design, that has been operating since 1957 at a power of 2 MW. Irradiated (spent) fuels have been stored at the facility during the various years of operation. At present there are 40 spent fuel assemblies at dry storage, 79 spent fuel assemblies at wet storage and 30 fuel assemblies in the core. The oldest fuels are of United States origin, made with U-Al alloy, both of LEU and HEU MTR fuel type. Many of these fuel assemblies have corrosion pits along their lateral fuel plates. These pits originate by galvanic corrosion between the fuel plate and the stainless steel storage racks. As a consequence of the possibility of sending the irradiated old fuels back the U.S.A., sipping tests were performed with the spent fuel assemblies. The reason for this was to evaluate their 137 Cs leaking rate, if any. This work describes the procedure and methodology used to perform the sipping tests with the fuel assemblies at the storage pool, and presents the results obtained for the 137 Cs sipping water activity for each fuel assembly. A correlation is made between the corrosion pits and the activity values measured. A 137 Cs leaking rate is determined and compared to the criteria established for canning spent fuel assemblies before shipment

  7. An Integrated Management System (IMS) for JM-1 SLOWPOKE-2 research reactor in Jamaica: experiences in documentation

    International Nuclear Information System (INIS)

    Warner, T.

    2014-01-01

    Since the first criticality in March 1984, the Jamaica SLOWPOKE-2 research reactor at the University of the West Indies, Mona located in the department of the International Centre for Environmental and Nuclear Sciences (ICENS) has operated for approximately 52% of the lifetime of the existing core configuration. The 20kW pool type research reactor has been primarily used for neutron activation analysis in environmental, agricultural, geochemical, health-related studies and mineral exploration in Jamaica. The involvement of the JM-1 reactor for research and teaching activities has segued into commercial applications which, coupled with the current core conversion programme from HEU to LEU, has demanded the implementation of management systems to satisfy regulatory requirements and assure compliance with internationally defined quality standards. At ICENS, documentation related to the Quality Management System aspect of an Integrated Management System (IMS) is well underway. The quality system will incorporate operational and nuclear safety, training, maintenance, design, utilization, occupational health and safety, quality service, and environmental management for its Nuclear Analytical Laboratory, NAL. The IMS is being designed to meet the requirements of the IAEA GS-R-3 with additional controls from international standards including: ISO/IEC 17025:2005, ISO 9001:2008, ISO 14001:2004 and OHSAS 18001:2007. This paper reports on the experiences of the documentation process in a low power reactor facility characterized by limited human resource, where innovative mechanisms of system automation and modeling are included to increase productivity and efficiency. (author)

  8. An Integrated Management System (IMS) for JM-1 SLOWPOKE-2 research reactor in Jamaica: experiences in documentation

    Energy Technology Data Exchange (ETDEWEB)

    Warner, T., E-mail: traceyann.warner02@uwimona.edu.jm [Univ. of West Indies, Mona (Jamaica)

    2014-07-01

    Since the first criticality in March 1984, the Jamaica SLOWPOKE-2 research reactor at the University of the West Indies, Mona located in the department of the International Centre for Environmental and Nuclear Sciences (ICENS) has operated for approximately 52% of the lifetime of the existing core configuration. The 20kW pool type research reactor has been primarily used for neutron activation analysis in environmental, agricultural, geochemical, health-related studies and mineral exploration in Jamaica. The involvement of the JM-1 reactor for research and teaching activities has segued into commercial applications which, coupled with the current core conversion programme from HEU to LEU, has demanded the implementation of management systems to satisfy regulatory requirements and assure compliance with internationally defined quality standards. At ICENS, documentation related to the Quality Management System aspect of an Integrated Management System (IMS) is well underway. The quality system will incorporate operational and nuclear safety, training, maintenance, design, utilization, occupational health and safety, quality service, and environmental management for its Nuclear Analytical Laboratory, NAL. The IMS is being designed to meet the requirements of the IAEA GS-R-3 with additional controls from international standards including: ISO/IEC 17025:2005, ISO 9001:2008, ISO 14001:2004 and OHSAS 18001:2007. This paper reports on the experiences of the documentation process in a low power reactor facility characterized by limited human resource, where innovative mechanisms of system automation and modeling are included to increase productivity and efficiency. (author)

  9. TRIGA reactor operating experience

    International Nuclear Information System (INIS)

    Anderson, T.V.

    1970-01-01

    The Oregon State TRIGA Reactor (OSTR) has been in operation 3 years. Last August it was upgraded from 250 kW to 1000 kW. This was accomplished with little difficulty. During the 3 years of operation no major problems have been experienced. Most of the problems have been minor in nature and easily corrected. They came from lazy susan (dry bearing), Westronics Recorder (dead spots in the range), The Reg Rod Magnet Lead-in Circuit (a new type lead-in wire that does not require the lead-in cord to coil during rod withdrawal hss been delivered, much better than the original) and other small corrections

  10. Reactor utilization, Part 1

    International Nuclear Information System (INIS)

    Martinc, R.; Stanic, A.

    1981-01-01

    The reactor operating plan for 1981 was subject to the needs of testing operation with the 80% enriched fuel and was fulfilled on the whole. This annex includes data about reactor operation, review of shorter interruptions due to demands of the experiments, data about safety shutdowns caused by power cuts. Period of operation at low power levels was used mostly for activation analyses, and the operation at higher power levels were used for testing and regular isotope production. Detailed data about samples activation are included as well as utilization of the reactor as neutron source and the operating plan for 1982 [sr

  11. Mark I 1/5-scale boiling water reactor pressure suppression experiment facility report

    International Nuclear Information System (INIS)

    Altes, R.G.; Pitts, J.H.; Ingraham, R.F.; Collins, E.K.; McCauley, E.W.

    1977-01-01

    An accurate Mark I 1 / 5 -scale, boiling water reactor (BWR), pressure suppression facility was designed and constructed at Lawrence Livermore Laboratory (LLL) in 11 months. Twenty-seven air tests using the facility are described. Cost was minimized by utilizing equipment borrowed from other LLL programs. The total value of borrowed equipment exceeded the program's budget of $2,020,000. Substantial flexibility in the facility was used to permit independent variation in the drywell pressure-time history, initial pressure in the drywell and toroidal wetwells, initial toroidal wetwell water level and downcomer length, vent line flow resistance, and vent line flow asymmetry. The two- and three-dimensional sectors of the toroidal wetwell provided significant data

  12. Postirradiation fracture toughness tests of ESR alloy HT-9 and modified 9Cr-1Mo alloy from UBR reactor experiments

    International Nuclear Information System (INIS)

    Hawthorne, J.R.; Reed, J.R.; Sprague, J.A.

    1984-01-01

    Alloy HT-9 and Modified 9Cr-1Mo are being evaluated for potential applications as first wall materials in magnetic fusion reactors. Objectives of the current research task were to test fatigue-precracked Charpy-V (PCC/sub v/) specimens from representative plates irradiated in the UBR reactor at 149 0 C or 300 0 C, and, to compare the results against postirradiation notch ductility data developed previously for the materials. Both plates represent electroslag refined (ESR) melt processing. PCC/sub v/ specimens of Alloy HT-9 and Modified 9Cr-1Mo alloy were irradiated at 300 0 C and 149 0 C, respectively, to approx.0.8 X 10 20 n/cm 2 , E > 0.1 MeV. During this period, postirradiation tests for fracture toughness were completed and results compared to notch ductility determinations from standard Charpy-V (C/sub v/) specimens irradiated in the same reactor experiments. Fracture surface examinations by SEM are also reported

  13. Detecting Dark Photons with Reactor Neutrino Experiments

    Science.gov (United States)

    Park, H. K.

    2017-08-01

    We propose to search for light U (1 ) dark photons, A', produced via kinetically mixing with ordinary photons via the Compton-like process, γ e-→A'e-, in a nuclear reactor and detected by their interactions with the material in the active volumes of reactor neutrino experiments. We derive 95% confidence-level upper limits on ɛ , the A'-γ mixing parameter, ɛ , for dark-photon masses below 1 MeV of ɛ reactors as potential sources of intense fluxes of low-mass dark photons.

  14. Neutron irradiation of V-Cr-Ti alloys in the BOR-60 fast reactor: Description of the fusion-1 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Rowcliffe, A.F. [Oak Ridge National Laboratory, TN (United States); Tsai, H.C.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-08-01

    The FUSION-1 irradiation capsule was inserted in Row 5 of the BOR-60 fast reactor in June 1995. The capsule contains a collaborative RF/U.S. experiment to investigate the irradiation performance of V-Cr-Ti alloys in the temperature range 310 to 350{degrees}C. This report describes the capsule layout, specimen fabrication history, and the detailed test matrix for the U.S. specimens. A description of the operating history and neutronics will be presented in the next semiannual report.

  15. As-Run Physics Analysis for the UCSB-1 Experiment in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nielsen, Joseph Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    The University of California Santa Barbara (UCSB) -1 experiment was irradiated in the A-10 position of the ATR. The experiment was irradiated during cycles 145A, 145B, 146A, and 146B. Capsule 6A was removed from the test train following Cycle 145A and replaced with Capsule 6B. This report documents the as-run physics analysis in support of Post-Irradiation Examination (PIE) of the test. This report documents the as-run fluence and displacements per atom (DPA) for each capsule of the experiment based on as-run operating history of the ATR. Average as-run heating rates for each capsule are also presented in this report to support the thermal analysis.

  16. Experience in utilizing research reactors in Yugoslavia

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J.; Raisic, N. [Boris Kidric Institute of Nuclear Sciences VINCA, Belgrade (Yugoslavia); Copic, M.; Gabrovsek, Z. [Jozef Stefan Institute Ljubljana (Yugoslavia)

    1972-07-01

    The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied by means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro

  17. Experience in utilizing research reactors in Yugoslavia

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.; Raisic, N.; Copic, M.; Gabrovsek, Z.

    1972-01-01

    The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied by means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro

  18. US graphite reactor D ampersand D experience

    International Nuclear Information System (INIS)

    Garrett, S.M.K.; Williams, N.C.

    1997-02-01

    This report describes the results of the U.S. Graphite Reactor Experience Task for the Decommissioning Strategy Plan for the Leningrad Nuclear Power Plant (NPP) Unit 1 Study. The work described in this report was performed by the Pacific Northwest National Laboratory (PNNL) for the Department of Energy (DOE)

  19. Computer simulations of a 1/5-scale experiment of a Mark I boiler water reactor pressure-suppression system under hypothetical LOCA conditions

    International Nuclear Information System (INIS)

    Edwards, L.L.

    1978-01-01

    The CHAMP computer code was employed to simulate a plane-geometry cross section of a Mark I boiling water reactor toroidal pressure suppression system air discharge experiment under hypothetical loss-of-coolant accident conditions. The experiments were performed at the Lawrence Livermore Laboratory on a 1 / 5 -scale model of the Peach Bottom Nuclear Power Plant

  20. Experiences with interpretation and application of the German Atomic Energy Act for the German research reactors in Geesthacht FRG-1 and FRG-2

    International Nuclear Information System (INIS)

    Krull, W.

    1980-01-01

    The German research reactors FRG-1 and FRG-2 have passed different types of licensing procedures in the past years. It is reported about the experiences we have got in the interpretation and application of section 7 of the German Atomic Energy Act. Following these experiences an estimation is done for the licensing procedure for the reduction of the uranium enrichment. (orig.) [de

  1. Irradiation experiments of recycled PuO2-UO2 fuels by SAXTON reactor, (1)

    International Nuclear Information System (INIS)

    Yumoto, Ryozo; Akutsu, Hideo

    1975-01-01

    Seventy two mixed oxide fuel rods made by PNC were irradiated in Saxton Core 3. This paper generally describes the fuel specifications, the power history of the fuel and the post-irradiation examination of the PNC fuel. The specifications of the 4.0 w/o and 5.0 w/o enriched PuO 2 fuel rods with zircaloy-4 cladding are presented in a table and a figure. The positions of PNC fuel rods in the Saxton reactor are shown in a figure. Sixty eight 5.0 w/o PuO 2 -UO 2 fuel rods were assembled in a 9 x 9 rod array together with zircaloy-4 bars, a flux thimble, and a Sb-Be source. The power history of the Saxton Core 3 and the irradiation history of the PNC fuel rods are summarized in tables. The peak power and burnup of each fuel rod and the axial power profile are also presented. The maximum linear power rate and burnup attained were 512W/cm and 8700 MWD/T, respectively. As for the post irradiation examination, the items of nondestructive test, destructive test, and cladding test are presented together with the working flow diagram of the examination. It is concluded that the performance of all fuel rods was safe and satisfactory throughout the power history. (Aoki, K.)

  2. PARR-2: reactor description and experiments

    International Nuclear Information System (INIS)

    Wyne, M.F.; Meghji, J.H.

    1990-12-01

    PARR-2 is a miniature neutron source reactor (MNSR) research reactor has been designed at the rate of 27 kW. Reactor assembly comprises of peaking characteristics with a self limiting flux. In this report reactor description with its assembly and instrumentation control system has been explained. The reactor engineering and physics experiments which can be performed on this reactor are explained in this report. PARR-2 is fueled with HEU fuel pins which are about 90% enriched in U-235. Specific requirements for the safety of the reactor, its building and the personnel, normal instrumentation as required in an industrial environment is sufficient. (A.B.)

  3. Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

    International Nuclear Information System (INIS)

    Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo

    2011-01-01

    IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

  4. Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo, E-mail: ahiru@ipen.b, E-mail: wricci@ipen.b, E-mail: carvalho@ipen.b, E-mail: jrretta@ipen.b, E-mail: amneto@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

  5. Neutron field for activation experiments in horizontal channel of training reactor VR-1

    Czech Academy of Sciences Publication Activity Database

    Štefánik, Milan; Katovsky, K.; Vinš, M.; Šoltéš, J.; Závorka, L.

    2014-01-01

    Roč. 104, NOV (2014), s. 302-305 ISSN 0969-806X. [1st International Conference on Dosimetry and its Applications (ICDA). Prague, 23.6.2013-28.6.2013] R&D Projects: GA MŠk LG14004 Institutional support: RVO:61389005 Keywords : spectral index * neutron spectrometry * dosimetry-foils activation technique * irradiation channel * reaction rate * Gamma -spectroscopy Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.380, year: 2014

  6. Final air test results for the 1/5-scale Mark I boiling water reactor pressure suppression experiment

    International Nuclear Information System (INIS)

    Collins, E.K.; Lai, W.

    1977-01-01

    A loss-of-coolant accident (LOCA) in a boiling-water reactor (BWR) power plant has never occurred. However, because this type of accident is particularly severe, it is used as a principal basis for design. During a hypothetical LOCA in a Mark I BWR, air followed by steam is injected from a drywell into a toroidal wetwell about half-filled with water. A series of consistent, versatile, and accurate air-water tests simulating LOCA conditions was completed in the Lawrence Livermore Laboratory 1/5-Scale Mark I BWR Pressure Suppression Experimental Facility. Results from this test series were used to quantify the vertical loading function and to study the associated fluid dynamic phenomena. Detailed histories of vertical loads on the wetwell are shown. In particular, variations of hydrodynamic-generated vertical loads with changes in drywell pressurization rate, downcomer submergence, and the vent-line loss coefficient are established. Initial drywell overpressure, which partially preclears the downcomers of water, substantially reduces the peak vertical loads. Scaling relationships, developed from dimensional analysis and verified by bench-top experiments, allow the 1/5-scale results to be applied to a full-scale BWR power plant. This analysis leads to dimensionless groupings which are invariant. These groupongs show that if water is used as the working fluid, the magnitude of the forces in a scaled facility is reduced by the cube of the scale factor; the time when these forces occur is reduced by the square root of the scale factor

  7. Daya bay reactor neutrino experiment

    International Nuclear Information System (INIS)

    Cao Jun

    2010-01-01

    Daya Bay Reactor Neutrino Experiment is a large international collaboration experiment under construction. The experiment aims to precisely determine the neutrino mixing angle θ 13 by detecting the neutrinos produced by the Daya Bay Nuclear Power Plant. θ 13 is one of two unknown fundamental parameters in neutrino mixing. Its magnitude is a roadmap of the future neutrino physics, and very likely related to the puzzle of missing antimatter in our universe. The precise measurement has very important physics significance. The detectors of Daya Bay is under construction now. The full operation is expected in 2011. Three years' data taking will reach the designed the precision, to determine sin 2 2θ 13 to better than 0.01. Daya Bay neutrino detector is an underground large nuclear detector of low background, low energy, and high precision. In this paper, the layout of the experiment, the design and fabrication progress of the detectors, and some highlighted nuclear detecting techniques developed in the detector R and D are introduced. (author)

  8. Lessons from early experience in reactor development

    International Nuclear Information System (INIS)

    Allen, W.

    1976-09-01

    This paper deals with several issues in U.S. reactor development and demonstration experience. The focus is on the period between 1946 and 1963 during which the Atomic Energy Commission (AEC) guided early reactor research and development (R and D) and conducted the Power Reactor Demonstration Program

  9. Experiment of IEA-R1 reactor core cooling by air convection after pool water loss accident

    International Nuclear Information System (INIS)

    Torres, Walmir Maximo; Baptista Filho, Benedito Dias

    2000-01-01

    This paper presents a study of a Emergency Core Cooling to be applied to the IEA-R1 reactor. This system must have the characteristics of passive action, with water spraying over the core, and feeding by gravity from elevated reservoirs. In the evaluation, this system must demonstrate that when the reservoirs are emptied, the core cooling must assure to be fulfilled by air natural convection. This work presents the results of temperature distribution in a test section with plates electrically heated simulation the heat generation conditions on the most heated reactor element

  10. Mark I 1/5-scale boiling water reactor pressure suppression experiment. Quick-look report for test numbers 1.0(a) and 1.0(b) performed on March 4 and 8, 1977

    International Nuclear Information System (INIS)

    McCauley, E.W.; Pitts, J.H.

    1977-01-01

    The experimental results obtained from pressure suppression experiment numbers 1.0(a) and 1.0(b) that were performed on the Lawrence Livermore Laboratory's 1 / 5 -scale boiling water reactor (BWR) Mark I pressure suppression experimental facility are summarized

  11. Training courses at VR-1 reactor

    International Nuclear Information System (INIS)

    Sklenka, L.; Kropik, M.

    2006-01-01

    This paper describes one of the main purposes of the VR-1 training reactor utilization - i.e. extensive educational program. The educational program is intended for the training of university students and selected nuclear power plant personnel. The training courses provide them experience in reactor and neutron physics, dosimetry, nuclear safety and operation of nuclear facilities. At present, the training course participants can go through more than 20 standard experimental exercises; particular exercises for special training can be prepared. Approximately 200 university students become familiar with the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. The VR-1 reactor takes also part in Eugene Wigner Course on Reactor Physics Experiments in the framework of European Nuclear Educational Network (ENEN) association. Recently, training courses for Bulgarian research reactor specialists supported by IAEA were carried out. An attractive program including demonstration of reactor operation is prepared also for high school students. Every year, more than 1500 high school students come to visit the reactor, as do many foreigner visitors. (author)

  12. Recent U.S. reactor operating experience

    International Nuclear Information System (INIS)

    Stello, V. Jr.

    1977-01-01

    A qualitative assessment of U.S. and foreign reactor operating experience is provided. Recent operating occurrences having potentially significant safety impacts on power operation are described. An evaluation of the seriousness of each of these issues and the plans for resolution is discussed. A quantitative report on U.S. reactor operational experience is included. The details of the NRC program for evaluating and applying operating reactor experience in the regulatory process is discussed. A review is made of the adequacy of operating reactor safety and environmental margins based on actual operating experience. The Regulatory response philosophy to operating reactor experiences is detailed. This discussion indicates the NRC emphasis on the importance of a balanced action plan to provide for the protection of public safety in the national interest

  13. Experience with valves for PHWR reactors

    International Nuclear Information System (INIS)

    Narayan, K.; Mhetre, S.G.

    1977-01-01

    Material specifications and inspection and testing requirements of the valves meant for use in nuclear reactors are mentioned. In the heavy water systems (both primary and moderator) of a PHWR type reactor, the valves used are gate valves, globe valves, diaphragm valves, butterfly valves, check valves and relief valves. Their locations and functions they perform in the Rajasthan Atomic Power Station Unit-1 are described. Experience with them is given. The major problems encountered with them have been : (1) leakage from the stem seals and body bonnet joint, (2) leakage due to failure of diaphragm and/or washout of the packing and (3) malfunctioning. Measures taken to solve these are discussed. Finally a mention has been made of improved versions of valves, namely, metal diaphragm valve and inverted relief valve. (M.G.B.)

  14. Photographic and video techniques used in the 1/5-scale Mark I boiling water reactor pressure suppression experiment

    International Nuclear Information System (INIS)

    Dixon, D.; Lord, D.

    1978-01-01

    The report provides a description of the techniques and equipment used for the photographic and video recordings of the air test series conducted on the 1/5 scale Mark I boiling water reactor (BWR) pressure suppression experimental facility at Lawrence Livermore Laboratory (LLL) between March 4, 1977, and May 12, 1977. Lighting and water filtering are discussed in the photographic system section and are also applicable to the video system. The appendices contain information from the photographic and video camera logs

  15. Postirradiation notch ductility tests of ESR alloy HT-9 and modified 9Cr-1Mo alloy from UBR reactor experiments

    International Nuclear Information System (INIS)

    Hawthorne, J.R.

    1984-01-01

    During this period, irradiation exposures at 300 0 C and 150 0 C to approx. 8 x 10 19 n/cm 2 , E > 0.1 MeV, were completed for the Alloy HT-9 plate and the modified Alloy 9Cr-1Mo plates, respectively. Postirradiation tests of Charpy-V (C/sub v/) specimens were completed for both alloys; other specimen types included in the reactor assemblies were fatigue precracked Charpy-V (PCC/sub v/), half-size Charpy-V, and in the case of the modified 9Cr-1Mo, 2.54 mm thick compact tension specimens

  16. International Experience with Fast Reactor Operation & Testing

    International Nuclear Information System (INIS)

    Sackett, John I.; Grandy, C.

    2013-01-01

    Conclusion: • Worldwide experience with fast reactors has demonstrated the robustness of the technology and it stands ready for worldwide deployment. • The lessons learned are many and there is danger that what has been learned will be forgotten given that there is little activity in fast reactor development at the present time. • For this reason it is essential that knowledge of fast reactor technology be preserved, an activity supported in the U.S. as well as other countries

  17. Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 4. Numerical analysis of 1/10 scaled water experiment with the AQUA code

    International Nuclear Information System (INIS)

    Muramatu, Toshiharu; Yamaguchi, Akira

    2004-01-01

    A large-scale sodium-cooled fast breeder reactor in the feasibility studies on commercialized fast reactors has a feature of consideration of thorough simplified and compacted systems and components design to realize drastic economical improvements. Therefore, special attentions should be paid to thermohydraulic designs for gas entrainment behavior from free surface, flow-induced vibration of in-vessel components, thermal stratification in the plenum, thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. A numerical analysis was carried out with a multi-dimensional code AQUA to confirm an applicability to the evaluations for the in-vessel thermohydraulic phenomena using a 1/10 scaled water experiment simulating the large-scale fast breeder reactor in the feasibility studies. From the analysis, the following results were obtained. (1) In-vessel thermohydraulics characterized by a radiated flow pattern to the reactor vessel wall and a strong upward flow through a slit of the upper core structures were evaluated. These characteristics agreed approximately with the water experiment. (2) The upward velocity values at the slit agreed well with the experimental data under a condition of γ z = 0.3 and ξ z = 0.5, though overall evaluations of the in-vessel thermohydraulics were failed to predict quantitatively. (3) The AQUA code is applicable to the in-vessel thermohydraulics evaluations in the feasibility studies, though it is necessary to make further modifications of the calculational models for accurate evaluations. On the one hand, it was confirmed that calculated results for the 1/10 water experimental model and the 1/1 actual-scaled model agreed quantitatively for the in-vessel thermohydraulics characteristics indicated above. (author)

  18. Reactor physics experiment plan using TCA

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, Shoichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-06-01

    The Reduced-Moderation Water Reactor (RMWR) is one of the next generation water-cooled reactors, which aims at effective utilization of uranium resource, high burn-up, long operation cycle, and plutonium multi-recycle. For verification of the feasibility, negative void reactivity coefficient and conversion ratio more than 1.0 must be confirmed. This report is to plan critical experiments using TCA in JAERI. Critical Experiments performed so far in Europe and Japan are reviewed, and no useful data are available for RMWR development. Critical experiments using TCA (Tank Type Critical Assembly) in JAERI are planned. MOX fuel rods should be prepared for the experiments and some modifications of equipment are needed for use of MOX fuel rods. This report describes the preliminary plan of physics experiments. The number of MOX-fuel rods used in the experiments is obtained by calculations and modification of the equipment for the experiments are shown. New MOX fuel and UO{sub 2} fuel rods are necessary for the RMWR critical experiments. Number of MOX fuel rods is 1000 for Plutonium fissile enrichment of 5 wt%, 1000 for 10 wt%, 1500 for 15 wt% and 500 for 20 wt%, respectively. Depleted UO{sub 2} fuel rods for blanket/buffer region are 4000. Driver fuel rods of 4.9 wt% UO{sub 2} are 3000. Modification of TCA facility is requested to treat the large amount of MOX fuels from safety point of view. Additional shielding device at the top of the tank for loading the MOX fuels and additional safety plates to ensure safety are requested. The core is divided into two regions by inserting an inner tank to avoid criticality in MOX region only. The test region is composed by MOX fuel rods in the inner tank. Criticality is established by UO{sub 2} driver fuel rods outside of the inner tank. (Tsuchihashi, K.)

  19. Experiments with the SUR 100 training reactor

    International Nuclear Information System (INIS)

    Milicic, B.

    1984-06-01

    This paper contains a compilation of various experiments using the SUR - 100 reactor for training purposes, which have been widly proved in practical work at the School for Nuclear Technology of the Karlsruhe Research Center. (orig.) [de

  20. Operational and reliability experience with reactor instrumentation

    International Nuclear Information System (INIS)

    Dixon, F.; Gow, R.S.

    1978-01-01

    In the last 15 years the CEGB has experienced progressive plant development, integration and changes in operating regime through nine nuclear (gas-cooled reactor) power stations with corresponding instrumentation advances leading towards more refined centralized control. Operation and reliability experience with reactor instrumentation is reported in this paper with reference to the progressive changes related to the early magnox, late magnox and AGR periods. Data on instrumentation reliability in terms of reactor forced outages are presented and show that the instrumentation contributions to loss of generating plant availability are small. Reactor safety circuits, neutron flux and temperature measurements, gas analysis and vibration monitoring are discussed. In reviewing the reactor instrumentation the emphasis is on reporting recent experience, particularly on AGR equipment, but overall performance and changes to magnox equipment are included so that some appreciation can be obtained of instrumentation requirements with respect to plant lifetimes. (author)

  1. Validation of the MC{sup 2}-3/DIF3D Code System for Control Rod Worth via the BFS-75-1 Reactor Physics Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this paper, control rod worths of the BFS-75-1 reactor physics experiments were examined using continuous energy MCNP models and deterministic MC2-3/DIF3D models based on the ENDF/B-VII.0 library. We can conclude that the ENDF/B-VII.0 library shows very good agreement in small-size metal uranium fuel loaded core which is surrounded by the depleted uranium blanket. However, the control rod heterogeneity effect reported by the reference is not significant in this problem because the tested control rod models were configured by single rod. Hence comparison with other control rod worth measurements data such as the BFS-109-2A reactor physics experiment is planned as a future study. The BFS-75-1 critical experiment was carried out in the BFS-1 facility of IPPE in Russia within the framework of validating an early phase of KALIMER- 150 design. The Monte-Carlo model of the BFS- 75-1 critical experiment had been developed. However, due to incomplete information for the BFS- 75-1 experiments, Monte-Carlo models had been generated for the reference criticality and sodium void reactivity measurements with disk-wise homogeneous model. Recently, KAERI performed another physics experiment, BFS-109-2A, by collaborating with Russian IPPE. During the review process of the experimental report of the BFS-109-2A critical experiments, valuable information for the BFS-1 facility which can also be used for the BFS-75-1 experiments was discovered.

  2. Reactor operator screening test experiences

    International Nuclear Information System (INIS)

    O'Brien, W.J.; Penkala, J.L.; Witzig, W.F.

    1976-01-01

    When it became apparent to Duquesne Light Company of Pittsburgh, Pennsylvania, that the throughput of their candidate selection-Phase I training-reactor operator certification sequence was something short of acceptable, the utility decided to ask consultants to make recommendations with respect to candidate selection procedures. The recommendation implemented was to create a Nuclear Training Test that would predict the success of a candidate in completing Phase I training and subsequently qualify for reactor operator certification. The mechanics involved in developing and calibrating the Nuclear Training Test are described. An arbitration decision that resulted when a number of International Brotherhood of Electrical Workers union employees filed a grievance alleging that the selection examination was unfair, invalid, not job related, inappropriate, and discriminatorily evaluated is also discussed. The arbitration decision favored the use of the Nuclear Training Test

  3. Experience from and research activities at the Otaniemi TRIGA reactor

    International Nuclear Information System (INIS)

    Bars, Bruno

    1976-01-01

    Experience from the Finnish TRIGA Reactor is reported, small changes and improvements in the control console of the Fir-1 reactor have been made. A minicomputer based data collecting system is planned and installed. It will be used for collecting data from operation and radiation monitors including the new isotope laboratory, and also simultaneously smaller experiments such as control rod calibration. A minicomputer is used for on-line reactor noise studies. The automatic uranium analyzer has a maximum sensitivity of 0.03 μg U 235 and 1.2 Th 232 . The system is now used at a sampling rate of about one sample per minute. (author)

  4. Results of the Nucifer reactor neutrino experiment

    Energy Technology Data Exchange (ETDEWEB)

    Buck, Christian; Lindner, Manfred [MPIK Heidelberg (Germany)

    2016-07-01

    Nuclear reactors are a strong and pure source of electron antineutrinos. With neutrino experiments close to compact reactor cores new insights into neutrino properties and reactor physics can be obtained. The Nucifer experiment is one of the pioneers in this class of very short baseline projects. Its detector to reactor distance is only about 7 m. The data obtained in the last years allowed to estimate the plutonium concentration in the reactor core by the neutrino flux measurement. This is of interest for safeguard applications and non proliferation efforts. The antineutrinos in Nucifer are detected via the inverse beta decay on free protons. Those Hydrogen nuclei are provided by 850 liters of organic liquid scintillator. For higher detection efficiency and background reduction the liquid is loaded with Gadolinium. Despite all shielding efforts and veto systems the background induced by the reactor activity and cosmogenic particles is still the main challenge in the experiment. The principle of the Nucifer detector is similar to the needs of upcoming experiments searching for sterile neutrinos. Therefore, the Nucifer results are also valuable input for the understanding and optimization of those next generation projects. The observation of sterile neutrinos would imply new physics beyond the standard model.

  5. Operational experience with Dragon reactor experiment of relevance to commercial reactors

    International Nuclear Information System (INIS)

    Capp, P.D.; Simon, R.A.

    1976-01-01

    An important part of the experience gained during the first ten years of successful power operation of the Dragon Reactor is relevant to the design and operation of future High Temperature Reactors (HTRs). The aspects presented in this paper have been chosen as being particularly applicable to larger HTR systems. Core performance under a variety of conditions is surveyed with particular emphasis on a technique developed for the identification and location of unpurged releasing fuel and the presence of activation and fission products in the core area. The lessons learned during the reflector block replacement are presented. Operating experience with the primary circuit identifies the lack of mixing of gas streams within the hot plenum and the problems of gas streaming in ducts. Helium leakage from the circuit is often greater than the optimum 0.1%/d. Virtually all the leakage problems are associated with the small bore instrument pipework essential for the many experiments associated with the Dragon Reactor Experiment (DRE). Primary circuit maintenance work confirms the generally clean state of the DRE circuit but identifies 137 Cs and 110 Agsup(m) as possible hazards if fuel emitting these isotopes is irradiated. (author)

  6. Study on reactor power transient characteristics (reactor training experiments). Control rod reactivity calibration by positive period method and other experiment

    International Nuclear Information System (INIS)

    Ozaki, Yoshihiko; Sunagawa, Takeyoshi

    2014-01-01

    In this paper, it is reported about some experiments that have been carried out in the reactor training that targets sophomore of the department of applied nuclear engineering, FUT. Reactor of Kinki University Atomic Energy Research Institute (UTR-KINKI) was used for reactor training. When each critical state was achieved at different reactor output respectively in reactor operating, it was confirmed that the control rod position at that time does not change. Further, control rod reactivity calibration experiments using positive Period method were carried out for shim safety rod and regulating rod, respectively. The results were obtained as reasonable values in comparison with the nominal value of the UTR-KINKI. The measurement of reactor power change after reactor scram was performed, and the presence of the delayed neutron precursor was confirmed by calculating the half-life. The spatial dose rate measurement experiment of neutrons and γ-rays in the reactor room in a reactor power 1W operating conditions were also performed. (author)

  7. Operating experience with the DRAGON High Temperature Reactor experiment

    International Nuclear Information System (INIS)

    Simon, R.A.; Capp, P.D.

    2002-01-01

    The Dragon Reactor Experiment in Winfrith/UK was a materials test facility for a number of HTR projects pursued in the sixties and seventies of the last century. It was built and managed as an OECD/NEA international joint undertaking. The reactor operated successfully between 1964 and 1975 to satisfy the growing demand for irradiation testing of fuels and fuel elements as well as for technological tests of components and materials. The paper describes the reactor's main experimental features and presents results of 11 years of reactor operation relevant for future HTRs. (author)

  8. Experience in operation of heavy water reactors

    International Nuclear Information System (INIS)

    Rotaru, Ion; Bilegan, Iosif; Ghitescu, Petre

    1999-01-01

    The paper presents the main topics of the CANDU owners group (COG) meeting held in Mangalia, Romania on 7-10 September 1998. These meetings are part of the IAEA program for exchange of information related mainly to CANDU reactor operation safety. The first meeting for PHWR reactors took place in Vienna in 1989, followed by those in Argentina (1991), India (1994) and Korea (1996). The topics discussed at the meeting in Romania were: operation experience and recent major events, performances of CANDU reactors and safe operation, nuclear safety and operation procedures of PHWR, programs and strategies of lifetime management of installations and components of NPPs, developments and updates

  9. Operational experience of the Marcoule reactors

    International Nuclear Information System (INIS)

    Conte, F.

    1963-01-01

    The results obtaining from three years operation of the reactors G-2, G-3 have made it possible to accumulate a considerable amount of operational experience of these reactors. The main original points: - the pre-stressed concrete casing - the possibility of loading while under power - automatic temperature control have been perfectly justified by the results of operation. The author confirms the importance of these original solutions and draws conclusions concerning the study of future nuclear power stations. (author) [fr

  10. Mechanistic Model for Ash Deposit Formation in Biomass Suspension Firing. Part 1: Model Verification by Use of Entrained Flow Reactor Experiments

    DEFF Research Database (Denmark)

    Hansen, Stine Broholm; Jensen, Peter Arendt; Jappe Frandsen, Flemming

    2017-01-01

    used to describe the deposit formation rates and deposit chemistry observed in a series of entrained flow reactor (EFR) experiments using straw and wood as fuels. It was found that model #1 was not able to describe the observed influence of temperature on the deposit buildup rates, predicting a much...... differ in the description of the sticking probability of impacted particles: model #1 employs a reference viscosity in the description of the sticking probability, while model #2 combines impaction of viscoelastic particles on a solid surface with particle capture by a viscous surface. Both models were...

  11. Present and future oscillation experiments at reactors

    International Nuclear Information System (INIS)

    Mikaehlyan, L.A.

    2001-01-01

    A report is presented on recent progress and developments (since the NANP'99 Conference) in the current and future long baseline (∼100 - 800 km) oscillation experiments at reactors. These experiments, under certain assumptions, can fully reconstruct the internal mass structure of the electron neutrino and provide a laboratory test of solar and atmospheric neutrino problems

  12. Extensive utilization of training reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, Karel; Sklenka, Lubomir

    2003-01-01

    Full text: The training reactor VR-1 Vrabec ('Sparrow'), operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly, it is designed and operated for training of students from Czech universities, preparing of experts for the Czech nuclear programme, as well as for certain research and development work, and for information programmes in the sphere of non-military nuclear energy use (public relation). The VR-1 training reactor is a pool-type light-water reactor based on enriched uranium with maximum thermal power 1kWth and short time period up to 5kW th . The moderator of neutrons is light demineralized water (H 2 O) that is also used as a reflector, a biological shielding, and a coolant. Heat is removed from the core with natural convection. The reactor core contains 14 to 18 fuel assemblies IRT-3M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The core is accommodated in a cylindrical stainless steel vessel - pool, which is filled with water. UR-70 control rods serve the reactor control and safe shutdown. Training of the VR-1 reactor provides students with experience in reactor and neutron physics, dosimetry, nuclear safety, and nuclear installation operation. Students from technical universities and from natural sciences universities come to the reactor for training. Approximately 200 university students are introduced to the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. Practical Course on Reactor Physics in Framework of European Nuclear Engineering Network has been newly introduced. Currently, students can try out more than 20 experimental exercises. Further training courses have been included

  13. Leak before break experience in CANDU reactors

    International Nuclear Information System (INIS)

    Price, E.G.; Moan, G.D.; Coleman, C.E.

    1988-04-01

    The paper describes how the requirements for Leak-Before-Break are met in CANDU reactors. The requirements are based on operational and laboratory experience. After the onset of leakage in a fuel channel from a delayed hydride crack, time is available to the operator to take action before the crack grows to an unstable length. The time available is calculated using different models which use crack growth data from small specimen tests. When the results from crack growth behaviour experiments, carried out on components removed from reactor are used in the model, the time available for operator response is about 100 hours

  14. CANDU reactors. Experience and innovation

    International Nuclear Information System (INIS)

    Hart, R.S.; Brooks, G.L.

    1989-02-01

    The title of this paper highlights two key considerations which must be properly balanced through good management in the evolution of any engineering product. Excessive reliance on experience will lead to product stagnation; excessive reliance on innovation will often lead to an unsatisfactory product, at least in the first generation of this product. To illustrate this balancing process, the paper reviews CANDU evolution and experience and the balance between proveness and innovation achieved through management of the evolution process from early prototypes to today's large-scale commercial units. A forecast of continuing evolutionary directions is included

  15. Candu reactors - experience and innovation

    International Nuclear Information System (INIS)

    Hart, R.S.; Brooks, G.L.

    1989-01-01

    The title of this paper highlights two key considerations which must be properly balanced through good management in the evolution of any engineering product. Excessive reliance on experience will lead to product stagnation; excessive reliance on innovation will often lead to an unsatisfactory product, at least in the first generation of this product. To illustrate this balancing process, the paper reviews CANDU evolution and experience and the balance between proveness and innovation achieved through management of the evolution process from early prototypes to today's large-scale commercial units. A forecast of continuing evolutionary directions is included

  16. State system experience with safeguarding power reactors

    International Nuclear Information System (INIS)

    Roehnsch, W.

    1982-01-01

    This session describes the development and operation of the State System of Accountancy and Control in the German Democratic Republic, and summarizes operating experience with safeguards at power reactor facilities. Overall organization and responsibilities, containment and surveillance measures, materials accounting, and inspection procedures will be outlined. Cooperation between the IAEA, State system, facility, and supplier authorities will also be addressed

  17. Critical experiments supporting close proximity water storage of power reactor fuel. Technical progress report, July 1, 1978-September 30, 1978

    Energy Technology Data Exchange (ETDEWEB)

    Baldwin, M.N.; Hoovler, G.S.; Eng, R.L.; Welfare, F.G.

    1978-11-01

    Experimental measurements are being taken on critical configurations of clusters of fuel rods mocking up LWR-type fuel elements in close proximity water storage. The results will serve to benchmark the computer codes used in designing nuclear power reactor fuel storage racks. KENO calculations of Cores I to VI are within two standard deviations of the measured k/sub eff/ values.

  18. Reactor operations at SAFARI-1

    International Nuclear Information System (INIS)

    Vlok, J.W.H.

    2003-01-01

    A vigorous commercial programme of isotope production and other radiation services has been followed by the SAFARI-1 research reactor over the past ten years - superimposed on the original purpose of the reactor to provide a basic tool for nuclear research, development and education to the country at an institutional level. A combination of the binding nature of the resulting contractual obligations and tighter regulatory control has demanded an equally vigorous programme of upgrading, replacement and renovation of many systems in order to improve the safety and reliability of the reactor. Not least among these changes is the more effective training and deployment of operations personnel that has been necessitated as the operational demands on the reactor evolved from five days per week to twenty four hours per day, seven days per week, with more than 300 days per year at full power. This paper briefly sketches the operational history of SAFARI-1 and then focuses on the training and structuring currently in place to meet the operational needs. There is a detailed step-by-step look at the operator?s career plan and pre-defined milestones. Shift work, especially the shift cycle, has a negative influence on the operator's career path development, especially due to his unavailability for training. Methods utilised to minimise this influence are presented. The increase of responsibilities regarding the operation of the reactor, ancillaries and experimental facilities as the operator progresses with his career are discussed. (author)

  19. Moving ring reactor 'Karin-1'

    International Nuclear Information System (INIS)

    1983-12-01

    The conceptual design of a moving ring reactor ''Karin-1'' has been carried out to advance fusion system design, to clarify the research and development problems, and to decide their priority. In order to attain these objectives, a D-T reactor with tritium breeding blanket is designed, a commercial reactor with net power output of 500 MWe is designed, the compatibility of plasma physics with fusion engineering is demonstrated, and some other guideline is indicated. A moving ring reactor is composed mainly of three parts. In the first formation section, a plasma ring is formed and heated up to ignition temperature. The plasma ring of compact torus is transported from the formation section through the next burning section to generate fusion power. Then the plasma ring moves into the last recovery section, and the energy and particles of the plasma ring are recovered. The outline of a moving ring reactor ''Karin-1'' is described. As a candidate material for the first wall, SiC was adopted to reduce the MHD effect and to minimize the interaction with neutrons and charged particles. The thin metal lining was applied to the SiC surface to solve the problem of the compatibility with lithium blanket. Plasma physics, the engineering aspect and the items of research and development are described. (Kako, I.)

  20. Physics experiments with the operating reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cullington, G R; King, D C

    1973-09-27

    Experimental techniques have been developed and used on Dragon to give consistent information on excess reactivity and shut down margin. The reactivity measurements have been correlated with the theoretical calculations and have led to improvements in the calculations. The methods used and the results obtained are accepted by the Safety Committee as sufficient evidence for compliance with the fuel loading safety rules. Although the reactor was not designed as an experimental facility, flux and dose measurements experiments have been successfully carried out. Mass flow and negative reactivity transient measurements have been carried out. These are valuable for demonstration of the flexibility of the reactor system and for giving confidence in theoretical calculations.

  1. Physics experiment on the Dragon reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, C.

    1974-10-15

    The paper describes a set of DRAGON experiments planned to measure burn-up effects in DRAGON irradiated fuel. Irradiated fuel elements from DRAGON are to be subjected to reactivity measurements in the HECTOR experimental reactor to infer the residual U235 content followed by isotopic analyses at CEA laboratories in 1975. Fast neutron damage to DRAGON graphite is compared to fast neutron dose measurements using Ni58 (n,p) Co58 activation wires in both DRAGON and the DIDO MTR. Gamma scanning of irradiated fuel elements are used to compare axial power profiles to those derived from two-dimensional and three-dimensional calculations of the DRAGON reactor.

  2. Controlling chemistry parameters in nuclear reactors and power plants, plant chemistry specification requirements and compliance - an overview of TAPS 1 and 2 experiences

    International Nuclear Information System (INIS)

    Ravindranath; Muralidharan, K.; Save, C.B.; Patil, D.P.

    2006-01-01

    Tarapur Atomic Power Station -TAPS 1 and 2 is a twin unit Boiling Water Reactor (BWR) Nuclear Power Plant commissioned in the year 1969. Both units are running with capacity factor of more than 90 % in their 20 th cycle of operation as on today. The 220 MWe units were derated to 160 MWe during 1984 consequent to isolation of Secondary Steam Generators (SSG) in the 10 th cycle of operation due to SSG tube leaks. This paper presents an overview of Plant Chemistry Control measures and experiences during the last 38 years of operation. The overall plant chemistry performance of TAPS 1 and 2 observed is very good; which is evident from the material condition of various systems reflected in QC and I reports, NDT and ISI reports. This is also supported by the fact that both Units are showing excellent performance continuously during recent years. (author)

  3. Sensitivity Evaluation of the Daily Thermal Predictions of the AGR-1 Experiment in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Grant Hawkes; James Sterbentz; John Maki

    2011-05-01

    A temperature sensitivity evaluation has been performed for the AGR-1 fuel experiment on an individual capsule. A series of cases were compared to a base case by varying different input parameters into the ABAQUS finite element thermal model. These input parameters were varied by ±10% to show the temperature sensitivity to each parameter. The most sensitive parameters are the outer control gap distance, heat rate in the fuel compacts, and neon gas fraction. Thermal conductivity of the compacts and graphite holder were in the middle of the list for sensitivity. The smallest effects were for the emissivities of the stainless steel, graphite, and thru tubes. Sensitivity calculations were also performed varying with fluence. These calculations showed a general temperature rise with an increase in fluence. This is a result of the thermal conductivity of the fuel compacts and graphite holder decreasing with fluence.

  4. Experience from replacement and check of thermocouples during reconstruction of in-reactor temperature measurements at Bohunice V-1 units 1 and 2

    International Nuclear Information System (INIS)

    Slanina, M.; Stanc, J.

    2001-01-01

    Replacement of thermocouples in the protection tube blocks was a key phase of the reconstruction of in-reactor temperature measurements at Bohunice V-1 with regard to the success, reliability and impact on safety of unit operation. The replacement consisted of reliable and safe withdrawal of 216 old thermocouples, their disposal and installation of new thermocouples into dry channels. In the material presented, this phase of reconstruction is described in details, with focus on the evaluation of replacement quality and check activities carried out at the new installed thermocouples. (Authors)

  5. The SM and MIR reactors operation experience

    International Nuclear Information System (INIS)

    Kuprienko, V.A.; Klinov, A.V.; Svyatkin, M.N.; Shamardin, V.K.

    1995-01-01

    The SM and MIR operation experience show that continuous work on the problem of ageing, in all its aspects, allows for prolongation of the research plant life cycle by several folds as compared to the initial project. The redesigned SM-3 reactor will operate for another 20 years. The similar result is expected from the MIR planned reconstruction which scope will be the topic of future presentations. (orig.)

  6. Reactors Project Delivery: The Value of Experiance

    International Nuclear Information System (INIS)

    Stosic, V. Zoran

    2014-01-01

    State of Affairs: Energy Potential and Density versus Environmental Load of different Energy Sources, Development of Fuel into Energy/Electricity Generation, Production Costs of Electricity, Contributions of Nuclear Energy to Security of Energy Supply, Recent Nuclear Development, Public Support growing again. Projects Status: Reactors under Construction, Different Projects Industrial Schemes, Projects Overview. The Value of Experience: Licensing, Standardization on Early Engineering Activities, Supply Chain and Manufacturing of Heavy Components, Installation, Procurement. (author)

  7. Background studies for the MINER Coherent Neutrino Scattering reactor experiment

    International Nuclear Information System (INIS)

    Agnolet, G.; Baker, W.; Barker, D.; Beck, R.; Carroll, T.J.; Cesar, J.; Cushman, P.; Dent, J.B.; De Rijck, S.; Dutta, B.; Flanagan, W.; Fritts, M.; Gao, Y.; Harris, H.R.; Hays, C.C.; Iyer, V.

    2017-01-01

    The proposed Mitchell Institute Neutrino Experiment at Reactor (MINER) experiment at the Nuclear Science Center at Texas A&M University will search for coherent elastic neutrino-nucleus scattering within close proximity (about 2 m) of a 1 MW TRIGA nuclear reactor core using low threshold, cryogenic germanium and silicon detectors. Given the Standard Model cross section of the scattering process and the proposed experimental proximity to the reactor, as many as 5–20 events/kg/day are expected. We discuss the status of preliminary measurements to characterize the main backgrounds for the proposed experiment. Both in situ measurements at the experimental site and simulations using the MCNP and GEANT4 codes are described. A strategy for monitoring backgrounds during data taking is briefly discussed.

  8. Background studies for the MINER Coherent Neutrino Scattering reactor experiment

    Energy Technology Data Exchange (ETDEWEB)

    Agnolet, G.; Baker, W. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Barker, D. [School of Physics & Astronomy, University of Minnesota, Minneapolis, MN 55455 (United States); Beck, R. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Carroll, T.J.; Cesar, J. [Department of Physics, University of Texas at Austin, Austin, TX 78712 (United States); Cushman, P. [School of Physics & Astronomy, University of Minnesota, Minneapolis, MN 55455 (United States); Dent, J.B. [Department of Physics, University of Louisiana at Lafayette, Lafayette, LA 70504 (United States); De Rijck, S. [Department of Physics, University of Texas at Austin, Austin, TX 78712 (United States); Dutta, B. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Flanagan, W. [Department of Physics, University of Texas at Austin, Austin, TX 78712 (United States); Fritts, M. [School of Physics & Astronomy, University of Minnesota, Minneapolis, MN 55455 (United States); Gao, Y. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Department of Physics & Astronomy, Wayne State University, Detroit 48201 (United States); Harris, H.R.; Hays, C.C. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Iyer, V. [School of Physical Sciences, National Institute of Science Education and Research, Jatni - 752050 (India); and others

    2017-05-01

    The proposed Mitchell Institute Neutrino Experiment at Reactor (MINER) experiment at the Nuclear Science Center at Texas A&M University will search for coherent elastic neutrino-nucleus scattering within close proximity (about 2 m) of a 1 MW TRIGA nuclear reactor core using low threshold, cryogenic germanium and silicon detectors. Given the Standard Model cross section of the scattering process and the proposed experimental proximity to the reactor, as many as 5–20 events/kg/day are expected. We discuss the status of preliminary measurements to characterize the main backgrounds for the proposed experiment. Both in situ measurements at the experimental site and simulations using the MCNP and GEANT4 codes are described. A strategy for monitoring backgrounds during data taking is briefly discussed.

  9. ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Kahler, A.C.; Herman, M.; Kahler,A.C.; MacFarlane,R.E.; Mosteller,R.D.; Kiedrowski,B.C.; Frankle,S.C.; Chadwick,M.B.; McKnight,R.D.; Lell,R.M.; Palmiotti,G.; Hiruta,H.; Herman,M.; Arcilla,R.; Mughabghab,S.F.; Sublet,J.C.; Trkov,A.; Trumbull,T.H.; Dunn,M.

    2011-12-01

    The ENDF/B-VII.1 library is the latest revision to the United States Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 423 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [M. B. Chadwick et al., 'ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data,' Nuclear Data Sheets, 112, 2887 (2011)]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unmoderated and uranium reflected {sup 235}U and {sup 239}Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also

  10. ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Kahler, A. [Los Alamos National Laboratory (LANL); Macfarlane, R E [Los Alamos National Laboratory (LANL); Mosteller, R D [Los Alamos National Laboratory (LANL); Kiedrowski, B C [Los Alamos National Laboratory (LANL); Frankle, S C [Los Alamos National Laboratory (LANL); Chadwick, M. B. [Los Alamos National Laboratory (LANL); Mcknight, R D [Argonne National Laboratory (ANL); Lell, R M [Argonne National Laboratory (ANL); Palmiotti, G [Idaho National Laboratory (INL); Hiruta, h [Idaho National Laboratory (INL); Herman, Micheal W [Brookhaven National Laboratory (BNL); Arcilla, r [Brookhaven National Laboratory (BNL); Mughabghab, S F [Brookhaven National Laboratory (BNL); Sublet, J C [Culham Science Center, Abington, UK; Trkov, A. [Jozef Stefan Institute, Slovenia; Trumbull, T H [Knolls Atomic Power Laboratory; Dunn, Michael E [ORNL

    2011-01-01

    The ENDF/B-VII.1 library is the latest revision to the United States' Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 423 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [1]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unrnoderated and uranium reflected (235)U and (239)Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also confirmed for selected actinide reaction rates such as (236)U; (238,242)Pu and (241,243)Am capture in fast systems. Other deficiencies, such as the overprediction of Pu solution system critical

  11. IRPhEP-handbook, International Handbook of Evaluated Reactor Physics Benchmark Experiments

    International Nuclear Information System (INIS)

    Sartori, Enrico; Blair Briggs, J.

    2008-01-01

    1 - Description: The purpose of the International Reactor Physics Experiment Evaluation Project (IRPhEP) is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. This work of the IRPhEP is formally documented in the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments,' a single source of verified and extensively peer-reviewed reactor physics benchmark measurements data. The IRPhE Handbook is available on DVD. You may request a DVD by completing the DVD Request Form available at: http://irphep.inl.gov/handbook/hbrequest.shtml The evaluation process entails the following steps: 1. Identify a comprehensive set of reactor physics experimental measurements data, 2. Evaluate the data and quantify overall uncertainties through various types of sensitivity analysis to the extent possible, verify the data by reviewing original and subsequently revised documentation, and by talking with the experimenters or individuals who are familiar with the experimental facility, 3. Compile the data into a standardized format, 4. Perform calculations of each experiment with standard reactor physics codes where it would add information, 5. Formally document the work into a single source of verified and peer reviewed reactor physics benchmark measurements data. The International Handbook of Evaluated Reactor Physics Benchmark Experiments contains reactor physics benchmark specifications that have been derived from experiments that were performed at various nuclear experimental facilities around the world. The benchmark specifications are intended for use by reactor physics personal to validate calculational techniques. The 2008 Edition of the International Handbook of Evaluated Reactor Physics Experiments contains data from 25 different

  12. Advanced test reactor testing experience-past, present and future

    International Nuclear Information System (INIS)

    Marshall, Frances M.

    2006-01-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors - US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, which places the capsule in direct contact with the primary coolant. The next level of experiment complexity is an instrumented lead experiment, which allows for active control of experiment conditions during the irradiation. The most complex experiment is the pressurized water loop, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans

  13. Ageing management experience at NUR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Melllal, Sabrina; Rezig, Mohamed; Zamoun, Rachid; Ameur, Azeddin [Nuclear Research Center of Draria, Algiers (Algeria)

    2013-07-01

    NUR is a 1 MW, open pool reactor moderated and cooled by light water. It was commissioned in 1989. NUR is used for education and training in Nuclear Engineering and related topics for COMENA and National Scientific Community. It is also used to perform R and D works and services at national and regional levels. In this presentation, we describe the methodology and the main development activities related to the ageing management at NUR reactor. These activities include inspection actions and development actions to introduce modifications, to solve obsolescence issues in view to implement the required preventive and curative maintenance programs and to improve the performances of the installation. These actions involved mainly the Operation Assistance System of the Reactor (OAS), the secondary cooling loop, the cooling tower. A new OAS using a new technology and having more possibilities than the older one was introduced in the control system of the reactor. The OAS hardware structure, software structure and the main functions performed are presented. The second loop is entirely refurbished. Two new cooling towers are installed and connected to the main heat exchanger with new piping and valves. The architecture of this new installation is described and the performance assessed. Other actions which involve auxiliary systems like emergency electrical system, air pneumatic system and automatic fire extinguishing are presented.

  14. Experience with mechanical segmentation of reactor internals

    International Nuclear Information System (INIS)

    Carlson, R.; Hedin, G.

    2003-01-01

    Operating experience from BWE:s world-wide has shown that many plants experience initial cracking of the reactor internals after approximately 20 to 25 years of service life. This ''mid-life crisis'', considering a plant design life of 40 years, is now being addressed by many utilities. Successful resolution of these issues should give many more years of trouble-free operation. Replacement of reactor internals could be, in many cases, the most favourable option to achieve this. The proactive strategy of many utilities to replace internals in a planned way is a market-driven effort to minimize the overall costs for power generation, including time spent for handling contingencies and unplanned outages. Based on technical analyses, knowledge about component market prices and in-house costs, a cost-effective, optimized strategy for inspection, mitigation and replacements can be implemented. Also decommissioning of nuclear plants has become a reality for many utilities as numerous plants worldwide are closed due to age and/or other reasons. These facts address a need for safe, fast and cost-effective methods for segmentation of internals. Westinghouse has over the last years developed methods for segmentation of internals and has also carried out successful segmentation projects. Our experience from the segmentation business for Nordic BWR:s is that the most important parameters to consider when choosing a method and equipment for a segmentation project are: - Safety, - Cost-effectiveness, - Cleanliness, - Reliability. (orig.)

  15. Further experience in simulation of rod drop experiments in the Loviisa and Mochovce reactors

    International Nuclear Information System (INIS)

    Siltanen, P.; Kaloinen, E.; Tanskanen, A.; Mattila, R.

    2001-01-01

    Simulations of reactor scram experiments using the 3-dimensional kinetics code HEXTRAN have been updated for the initial cores of Loviisa-1 and 2 Mochovce-1 and have been extended to burned cores of Loviisa-1. In these simulations, the entire experiment is simulated dynamically, including the behaviour of the core, the signal of the ionization chamber, and the inverse point kinetics of the reactivity meter. The predicted output of the reactivity meter is compared with the output observed during the experiment (Authors)

  16. Basic experiments of reactor physics using the critical assembly TCA

    International Nuclear Information System (INIS)

    Obara, Toru; Igashira, Masayuki; Sekimoto, Hiroshi; Nakajima, Ken; Suzaki, Takenori.

    1994-02-01

    This report is based on lectures given to graduate students of Tokyo Institute of Technology. It covers educational experiments conducted with the Tank-Type Critical Assembly (TCA) at Japan Atomic Energy Research Institute in July, 1993. During this period, the following basic experiments on reactor physics were performed: (1) Critical approach experiment, (2) Measurement of neutron flux distribution, (3) Measurement of power distribution, (4) Measurement of fuel rod worth distribution, (5) Measurement of safety sheet worth by the rod drop method. The principle of experiments, experimental procedure, and analysis of results are described in this report. (author)

  17. Operation and maintenance of 1MW PUSPATI TRIGA reactor

    International Nuclear Information System (INIS)

    Adnan Bokhari; Mohammad Suhaimi Kassim

    2006-01-01

    The Malaysian Research Reactor, Reactor TRIGA PUSPATI (RTP) has been successfully operated for 22 years for various experiments. Since its commissioning in June 1982 until December 2004, the 1MW pool-type reactor has accumulated more than 21143 hours of operation, corresponding to cumulative thermal energy release of about 14083 MW-hours. The reactor is currently in operation and normally operates on demand, which is normally up to 6 hours a day. Presently the reactor core is made up of standard TRIAGA fuel element consists of 8.5 wt%, 12 wt% and 20 wt% types; 20%-enriched and stainless steel clad. Several measures such as routine preventive maintenance and improving the reactor support systems have been taken toward achieving this long successful operation. Besides normal routine utilization like other TRIGA reactors, new strategies are implemented for effective increase in utilization. (author)

  18. Plant experience of experimental fast reactor 'Joyo'

    International Nuclear Information System (INIS)

    1982-01-01

    The experimental fast reactor ''JOYO'' installed in Power Reactor and Nuclear Fuel Development Corp. (PNC) of Japan completed its operation using the first core (called MK-I core) in December, 1981, and the works to transfer to MK-2 core have been performed since January, 1982. In this report, the experiences obtained through the construction, test and operation of ''JOYO'' over 12 years from the start of erection in 1970 to the termination of operation in 1981 are described. The contents of the report are divided into design, construction, the outline of facilities, testing, operating and maintenance experiences, and the topics on MK-I operation. As for the construction, the design changes performed before the start of manufacture or construction and the improvement and trouble restoring works implemented at the start of overall functional tests are reported. As for testing, overall functional tests, criticality test, low power test and power increasing test are described in detail. The number of test items of overall functional testing reached 266. The rated output operation of the reactor at 75 MW was performed six times in 1980 and 1981 until the termination of operation. No fuel failure was detected in MK-I operation, and the stable operation performance of the FBR was proved through MK-I operation. The topics on the MK-I operation includes natural circulation test, the measurement of total leakage rate for the containment vessel, and wear-marks which are the trace of wear due to the contact of fuel pins with the wires wound around the adjacent fuel pins, found in the post irradiation examination of fuel. (Wakatsuki, Y.)

  19. The double chooz reactor neutrino experiment

    Energy Technology Data Exchange (ETDEWEB)

    Botella, I Gil [CIEMAT, Basic Research Department, Avenida Complutense, 22, 28040 Madrid (Spain)], E-mail: ines.gil@ciemat.es

    2008-05-15

    The Double Chooz reactor neutrino experiment will be the next detector to search for a non vanishing {theta}{sub 13} mixing angle with unprecedented sensitivity, which might open the way to unveiling CP violation in the leptonic sector. The measurement of this angle will be based in a precise comparison of the antineutrino spectrum at two identical detectors located at different distances from the Chooz nuclear reactor cores in France. Double Chooz is particularly attractive because of its capability to measure sin{sup 2} (2{theta}{sub 13}) to 3{sigma} if sin{sup 2}(2{theta}{sub 13}) > 0.05 or to exclude sin{sup 2}(2{theta}{sub 13}) down to 0.03 at 90% C.L. for {delta}m{sup 2} = 2.5 x 10{sup -3} eV{sup 2} in three years of data taking with both detectors. The construction of the far detector starts in 2008 and the first neutrino results are expected in 2009. The current status of the experiment, its physics potential and design and expected performance of the detector are reviewed.

  20. Intercomparison of liquid metal fast reactor seismic analysis codes. V.1: Validation of seismic analysis codes using reactor core experiments. Proceedings of a research co-ordination meeting held in Vienna, 16-17 November 1993

    International Nuclear Information System (INIS)

    1995-05-01

    The Research Co-ordination Meeting held in Vienna, 16-17 November 1993, was attended by participants from France, India, Italy, Japan and the Russian Federation. The meeting was held to discuss and compare the results obtained by various organizations for the analysis of Italian tests on PEC mock-up. The background paper by A. Martelli, et al., Italy, entitled Fluid-Structure Interaction Experiments of PEC Core Mock-ups and Numerical Analysis Performed by ENEA presented details on the Italian PEC (Prova Elementi di Combustibile, i.e. Fuel Element Test Facility) test data for the benchmark. Several papers were presented on the analytical investigations of the PEC reactor core experiments. The paper by M. Morishita, Japan, entitled Seismic Response Analysis of PEC Reactor Core Mock-up, gives a brief review of the Japanese data on the Monju mock-up core experiment which had been distributed to the participating countries through the IAEA. Refs, figs and tabs

  1. Advanced Reactor Fuels Irradiation Experiment Design Objectives

    International Nuclear Information System (INIS)

    Chichester, Heather Jean MacLean; Hayes, Steven Lowe; Dempsey, Douglas; Harp, Jason Michael

    2016-01-01

    This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

  2. Advanced Reactor Fuels Irradiation Experiment Design Objectives

    Energy Technology Data Exchange (ETDEWEB)

    Chichester, Heather Jean MacLean [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hayes, Steven Lowe [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dempsey, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

  3. Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

    Energy Technology Data Exchange (ETDEWEB)

    Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

    2016-06-15

    Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

  4. Neutron irradiation experiments for fusion reactor materials through JUPITER program

    International Nuclear Information System (INIS)

    Abe, K.; Namba, C.; Wiffen, F.W.; Jones, R.H.

    1998-01-01

    A Japan-USA program of irradiation experiments for fusion research, ''JUPITER'', has been established as a 6 year program from 1995 to 2000. The goal is to study ''the dynamic behavior of fusion reactor materials and their response to variable and complex irradiation environment''. This is phase-three of the collaborative program, which follows RTNS-II program (phase-1: 1982-1986) and FFTF/MOTA program (phase-2: 1987-1994). This program is to provide a scientific basis for application of materials performance data, generated by fission reactor experiments, to anticipated fusion environments. Following the systematic study on cumulative irradiation effects, done through FFTF/MOTA program. JUPITER is emphasizing the importance of dynamic irradiation effects on materials performance in fusion systems. The irradiation experiments in this program include low activation structural materials, functional ceramics and other innovative materials. The experimental data are analyzed by theoretical modeling and computer simulation to integrate the above effects. (orig.)

  5. Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

    International Nuclear Information System (INIS)

    Henry, A.F.

    1980-01-01

    Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

  6. Some scoping experiments for a space reactor

    International Nuclear Information System (INIS)

    Alexander, C.A.; Ogden, J.S.

    1983-01-01

    Some scoping experiments were performed to evaluate fuel performance in a lithium heat pipe reactor operating at a nominal 1500K heat pipe temperature. Fuel-coolant and fuel-coolant-clad relationships showed that once a failed heat pipe occurs temperatures can rise high enough so that large concentrations of uranium can be transported by the vapor phase. Upon condensation this uranium would be capable of penetrating heat pipes adjacent to the failed pipe. The potential for propagation of failure exists with UO 2 and a lithium heat pipe. Changing the composition of the metal of the heat pipe would have only a second order effect on the kinetics of the failure mechanism. Uranium carbide and nitride were considered as potential fuels which are nonreactive in a lithium environment. At high temperatures the nitride would be favored because of its better compatibility with potential cladding materials. Compositions of UN with small additions of YN appear to offer very attractive properties for a compact high temperature high power density reactor

  7. Storage experience in Hungary with fuel from research reactors

    International Nuclear Information System (INIS)

    Gado, J.; Hargitai, T.

    1996-01-01

    In Hungary several critical assemblies, a training reactor and a research reactor have been in operation. The fuel used in the research and training reactors are of Soviet origin. Though spent fuel storage experience is fairly good, medium and long term storage solutions are needed. (author)

  8. Channel type reactors with supercritical water coolant. Russian experience

    International Nuclear Information System (INIS)

    Kuznetsov, Y.N.; Gabaraev, B.A.

    2003-01-01

    Transition to coolant of supercritical parameters allows for principle engineering-andeconomic characteristics of light-water nuclear power reactors to be substantially enhanced. Russian experience in development of channel-type reactors with supercritical water coolant has demonstrated advantages and practical feasibility of such reactors. (author)

  9. Lessons from feedback of safety operating experience for reactor physics

    International Nuclear Information System (INIS)

    Suchomel, J.; Rapavy, S.

    1999-01-01

    Analyses of events in WWER operations as a part of safety experience feedback provide a valuable source of lessons for reactor physics. Examples of events from Bohunice operation will be shown such as events with inadequate approach to criticality, positive reactivity insertions, expulsion of a control rod from shut-down reactor, problems with reactor protection system and control rods. (Authors)

  10. Advances in Reactor Physics, Mathematics and Computation. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume one, are divided into 6 sessions bearing on: - session 1: Advances in computational methods including utilization of parallel processing and vectorization (7 conferences) - session 2: Fast, epithermal, reactor physics, calculation, versus measurements (9 conferences) - session 3: New fast and thermal reactor designs (9 conferences) - session 4: Thermal radiation and charged particles transport (7 conferences) - session 5: Super computers (7 conferences) - session 6: Thermal reactor design, validation and operating experience (8 conferences).

  11. Data acquisition. GRAAL experiment. Hybrid reactor experiment. AMS experiment

    International Nuclear Information System (INIS)

    Barancourt, D.; Barbier, G.; Bosson, G.; Bouvier, J.; Gallin-Martel, L.; Meillon, B.; Stassi, P.; Tournier, M.

    1997-01-01

    The main activity of the data acquisition team has consisted in hardware and software developments for the GRAAL experiment with the trigger board, for the 'Reacteurs Hybrides' group with an acquisition board ADCVME8V and for the AMS experiment with the monitoring of the aerogel detector. (authors)

  12. Backflushable filter experience at the N Reactor

    International Nuclear Information System (INIS)

    Ball, B.; Best, W.T.; Keith, R.C.

    1987-01-01

    The N Reactor is an 4000 MWt, light-water cooled, graphite-moderated reactor located on the Hanford Site in Washington State. A radwaste pilot plant to process plant effluent was constructed in order to maximize future efficiency when a full size radioactive processing facility is built. The pilot plant's purpose is to vary operational parameters such as filtration and ion exchange on a smaller scale to gather as much data as possible. The input to the pilot plant is radioactive drain lines from the N Reactor. The effluent passes through a backflushable filter and a series of ion exchange columns all scaled down from the future proposed facility. A backflushable filter was selected for this application because of the specific characteristics of the plant effluent and the potential reduced operating costs. The filter performance has been excellent in terms of filtration of the effluent. Typical total suspended solids in the plant effluent range from 1 to 6.1 ppm; the filter reduces this value to less than 0.1 ppm. In addition to outstanding filtration efficiency, the use of a precoat material on the filter has resulted in impressive decontamination factors. The filter has been successful in removing up to 50% of the influent activity. An improved performance of several nuclides over other filtration systems has also been achieved. By varying the composition and amount of precoat material on the filter, substantial reductions in waste volumes (and associated operating and disposal costs) have been demonstrated while maintaining a high degree of removal of both activity and total suspended solids

  13. Annual report on JEN-1 reactor

    International Nuclear Information System (INIS)

    Montes, J.

    1972-01-01

    In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

  14. Probabilistic risk analysis of Angra-1 reactor

    International Nuclear Information System (INIS)

    Spivak, R.C.; Collussi, I.; Silva, M.C. da; Onusic Junior, J.

    1986-01-01

    The first phase of probabilistic study for safety analysis and operational analysis of Angra-1 reactor is presented. The study objectives and uses are: to support decisions about safety problems; to identify operational and/or project failures; to amplify operator qualification tests to include accidents in addition to project base; to provide informations to be used in development and/or review of operation procedures in emergency, test and maintenance procedures; to obtain experience for data collection about abnormal accurences; utilization of study results for training operators; and training of evaluation and reliability techniques for the personnel of CNEN and FURNAS. (M.C.K.) [pt

  15. Technical Information on the Carbonation of the EBR-II Reactor, Summary Report Part 1: Laboratory Experiments and Application to EBR-II Secondary Sodium System

    Energy Technology Data Exchange (ETDEWEB)

    Steven R. Sherman

    2005-04-01

    Residual sodium is defined as sodium metal that remains behind in pipes, vessels, and tanks after the bulk sodium metal has been melted and drained from such components. The residual sodium has the same chemical properties as bulk sodium, and differs from bulk sodium only in the thickness of the sodium deposit. Typically, sodium is considered residual when the thickness of the deposit is less than 5-6 cm. This residual sodium must be removed or deactivated when a pipe, vessel, system, or entire reactor is permanently taken out of service, in order to make the component or system safer and/or to comply with decommissioning regulations. As an alternative to the established residual sodium deactivation techniques (steam-and-nitrogen, wet vapor nitrogen, etc.), a technique involving the use of moisture and carbon dioxide has been developed. With this technique, sodium metal is converted into sodium bicarbonate by reacting it with humid carbon dioxide. Hydrogen is emitted as a by-product. This technique was first developed in the laboratory by exposing sodium samples to humidified carbon dioxide under controlled conditions, and then demonstrated on a larger scale by treating residual sodium within the Experimental Breeder Reactor II (EBR-II) secondary cooling system, followed by the primary cooling system, respectively. The EBR-II facility is located at the Idaho National Laboratory (INL) in southeastern Idaho, U.S.A. This report is Part 1 of a two-part report. It is divided into three sections. The first section describes the chemistry of carbon dioxide-water-sodium reactions. The second section covers the laboratory experiments that were conducted in order to develop the residual sodium deactivation process. The third section discusses the application of the deactivation process to the treatment of residual sodium within the EBR-II secondary sodium cooling system. Part 2 of the report, under separate cover, describes the application of the technique to residual sodium

  16. Use of the VR-1 ''Vrabec'' training reactor

    International Nuclear Information System (INIS)

    Matejka, K.; Kolros, A.; Krops, S.; Polach, S.; Sklenka, L.

    1994-01-01

    An overview is presented of the extent and ways of using the VR-1 training reactor, which is operated by the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague. A list and the characteristics of 16 problems developed for teaching purposes is given, and the 14 faculties and 2 research institutes participating in the teaching activities are listed. The reactor is used in the education and training of nuclear scientists and engineers. The instrumentation, experimental, handling and operating tools, as well as documentation and texts relating to the reactor are described. The following examples of the teaching activities are included: a guided visit to the operating reactor site, reactor dynamics study and delayed neutron measurement, training course, and the basic criticality experiment. Nuclear safety aspects (hypothetical accidents, quality control and system qualification demonstration, safety culture) are stressed during the education. The reactor department is involved in international cooperation projects. (J.B.). 3 refs

  17. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    Energy Technology Data Exchange (ETDEWEB)

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)

    1994-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

  18. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    International Nuclear Information System (INIS)

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K.

    1994-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results

  19. The radon monitoring system in Daya Bay Reactor Neutrino Experiment

    International Nuclear Information System (INIS)

    Chu, M.C.; Kwan, K.K.; Kwok, M.W.; Kwok, T.; Leung, J.K.C.; Leung, K.Y.; Lin, Y.C.; Luk, K.B.; Pun, C.S.J.

    2016-01-01

    We developed a highly sensitive, reliable and portable automatic system (H 3 ) to monitor the radon concentration of the underground experimental halls of the Daya Bay Reactor Neutrino Experiment. H 3 is able to measure radon concentration with a statistical error less than 10% in a 1-h measurement of dehumidified air (R.H. 5% at 25 °C) with radon concentration as low as 50 Bq/m 3 . This is achieved by using a large radon progeny collection chamber, semiconductor α-particle detector with high energy resolution, improved electronics and software. The integrated radon monitoring system is highly customizable to operate in different run modes at scheduled times and can be controlled remotely to sample radon in ambient air or in water from the water pools where the antineutrino detectors are being housed. The radon monitoring system has been running in the three experimental halls of the Daya Bay Reactor Neutrino Experiment since November 2013.

  20. Evaporation rate measurement in the pool of IEAR-1 reactor

    International Nuclear Information System (INIS)

    Torres, Walmir Maximo; Cegalla, Miriam A.; Baptista Filho, Benedito Dias

    2000-01-01

    The surface water evaporation in pool type reactors affects the ventilation system operation and the ambient conditions and dose rates in the operation room. This paper shows the results of evaporation rate experiment in the pool of IEA-R1 research reactor. The experiment is based on the demineralized water mass variation inside cylindrical metallic recipients during a time interval. Other parameters were measured, such as: barometric pressure, relative humidity, environmental temperature, water temperature inside the recipients and water temperature in the reactor pool. The pool level variation due to water contraction/expansion was calculated. (author)

  1. Experience with reactor assembly of FBTR

    International Nuclear Information System (INIS)

    Srinivasan, G.; Ravishankar, K.; Babu, A.; Varadarajan, S.; Arumugam, P.; Sekhar, P.

    2006-01-01

    Reactor Assembly, also called Block Pile, is the heart of FBTR and houses the core, top and lateral shields, control rod drive mechanisms (CRDM), sodium inlet pipe and outlet pipes etc. Two major problems which arose during commissioning were reactor vessel tilt due to convection in cover gas space and failure of inflatable seals. The reactor vessel tilt was solved by Helium injection. Reactor was operated without pressurising the inflatable seals till 2005, when the seals were replaced. Other major problems in the course of twenty years of reactor operation were failure of three CRDM lower parts, Core Cover plate which houses the core thermocouples getting stuck in the fuel handling position, water leaks from the Biological Shield Cooling (BSC) coils around the reactor, failure of core wires in the trailing cables during fuel handling etc. This paper addresses the major problems faced and modifications carried out. (author)

  2. New safety experiments in decommissioned superheated steam reactor at Karlstein

    International Nuclear Information System (INIS)

    Koerting, K.

    1986-01-01

    This article gives a concise summary of the Status Report of the Superheated Steam Reactor Safety Program (PHDR) Project, held at KfK on Dec. 5, 1985. The results discussed dealt with fire experiments, shock tests simulating airplane crashes, temperature shocks in the reactor pressure vessel, studies of crack detection in pressure vessels and blasting experiments associated with nuclear plant decommissioning

  3. RA Research reactor Annual report 1981 - Part 1, Operation, maintenance and utilization of the RA reactor

    International Nuclear Information System (INIS)

    Sotic, O.; Milosevic, M.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.

    1981-12-01

    The RA nuclear reactor stopped operation after March 1979 campaign due to appearance of aluminium oxyhydrates deposits on the surface of fuel element claddings. Relevant decisions of the Sanitary inspection body of the Ministry of health and the Director General of the 'Boris Kidric' Institute of nuclear sciences, Vinca, banned further reactor operation until reasons caused aluminium oxyhydrates deposition are investigated and removed to enable regular reactor operation. Until the end of 1979 and during 1980, after a series of analyses and findings that caused cease of reactor operation, all the preparatory actions needed for restart were performed. Due to the fact that there is no emergency cooling system and no appropriate filtering system at the reactor, and according to the new regulations about start up of nuclear facilities, the Sanitary inspection body made a decision about temporary licence for reactor start-up meaning performance of the 'zero experiment' limiting the operating power to 1% of the nominal power. Accordingly the reactor was restarted on January 21 1981. Criticality was reached with the core made of 80% enriched fuel elements only. After the experiment was finished by the end of March a permission was demanded for operation at higher power levels at full power. Taking into account the state of the reactor components the operating licence was issued limiting the power to 2 MW until reconstruction of the ventilation system and construction of the emergency cooling system are fulfilled. Program of testing operation started on September 15 1981 increasing gradually the operating power. Thus the reactor was operated at 2 MW power for 15 days during November and December. The total production achieved in 1981 was 1698 MWh. This enabled isotopes production at the reactor during last two months. Control and maintenance of the reactor components and systems was done regularly and efficiently within limits imposed by availability of spare parts. The

  4. Experience and prospects for developing research reactors of different types

    International Nuclear Information System (INIS)

    Kuatbekov, R.P.; Tretyakov, I.T.; Romanov, N.V.; Lukasevich, I.B.

    2015-01-01

    NIKIET has a 60-year experience in the development of research reactors. Altogether, there have been more than 25 NIKIET-designed plants of different types built in Russia and 20 more in other countries, including pool-type water-cooled and water moderated research reactors, tank-type and pressure-tube research reactors, pressurized high-flux, heavy-water, pulsed and other research reactors. Most of the research reactors were designed as multipurpose plants for operation at research centers in a broad range of applications. Besides, unique research reactors were developed for specific application fields. Apart from the experience in the development of research reactor designs and the participation in the reactor construction, a unique amount of knowledge has been gained on the operation of research reactors. This makes it possible to use highly reliable technical solutions in the designs of new research reactors to ensure increased safety, greater economic efficiency and maintainability of the reactor systems. A multipurpose pool-type research reactor of a new generation is planned to be built at the Center for Nuclear Energy Science & Technology (CNEST) in the Socialist Republic of Vietnam to be used to support a spectrum of research activities, training of skilled personnel for Vietnam nuclear industry and efficient production of isotopes. It is exactly the applications a research reactor is designed for that defines the reactor type, design and capacity, and the selection of fuel and components subject to all requirements of industry regulations. The design of the new research reactor has a great potential in terms of upgrading and installation of extra experimental devices. (author)

  5. Proposed Reactor Operating Experience Feedback System Development

    International Nuclear Information System (INIS)

    Ahn, Seung Hoon; Kim, Min Chul; Huh, Chang Wook; Lee, Durk Hun; Bae, Koo Hyun

    2006-01-01

    Most events occurring in nuclear power plants are not individually significant, and prevented from progressing to accident conditions by a series of barriers against core damage and radioactive releases. Significant events, if occur, are almost always a breach of these multiple barriers. As illustrated in the 'Swiss cheese' model, the individual layers of defense or 'cheese slices' have weakness or 'holes.' These weaknesses are inconstant, i.e., the holes are open or close at random. When by chance all the holes are aligned, a hazard causes the significant event of concern. Elements of low significant events, inattention to detail, time or economic pressure, uncorrected poor practices/habits, marginal maintenance and equipment care, etc., make holes in the layers of defense; some elements may make more holes in different layers, incurring more chances to be aligned. An effective reduction of the holes, therefore, is gained through better knowledge or awareness of increasing trends of the event elements, followed by appropriate actions. According to the Swiss cheese metaphor, attention to the Operating Experience (OE) feedback system, as opposed to the individual and to randomness, is drawn from a viewpoint of reactor safety

  6. Reactor protection system. Revision 1

    International Nuclear Information System (INIS)

    Fairbrother, D.B.; Vincent, D.R.; Lesniak, L.M.

    1975-04-01

    The reactor protection system-II (RPS-II) designed for use on Babcock and Wilcox 145- and 205-fuel assembly pressurized water reactors is described. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W. (U.S.)

  7. Discussion of the use of the Dragon reactor as a facility for integral reactor physics experiments

    Energy Technology Data Exchange (ETDEWEB)

    Gutmann, H

    1972-06-05

    The purpose and use of the Dragon Reactor Experiment (DRE) has changed considerably during the years of its operation. The original purpose was to show that the principle of a High Temperature Reactor is sound and demonstrate its operation. After this achievement, the purpose of the Dragon reactor changed to the use as a fuel testing facility. During recent years, a new use of the DRE has been added to its use as a fuel testing facility, namely Fuel Element Design Testing. The current report covers reactor physics experiments aspects.

  8. The experiences of research reactor accident to safety improvement

    International Nuclear Information System (INIS)

    Wiranto, S.

    1999-01-01

    The safety of reactor operation is the main factor in order that the nuclear technology development program can be held according the expected target. Several experience with research reactor incidents must be learned and understood by the nuclear program personnel, especially for operators and supervisors of RSG-GA. Siwabessy. From the incident experience of research reactor in the world, which mentioned in the book 'Experience with research reactor incidents' by IAEA, 1995, was concluded that the main cause of research reactor accidents is understandless about the safety culture by the nuclear installation personnel. With learn, understand and compare between this experiences and the condition of RSG GA Siwabessy is expended the operators and supervisors more attention about the safety culture, so that RSG GA Siwabessy can be operated successfull, safely according the expected target

  9. The reactor antineutrino anomaly and low energy threshold neutrino experiments

    Science.gov (United States)

    Cañas, B. C.; Garcés, E. A.; Miranda, O. G.; Parada, A.

    2018-01-01

    Short distance reactor antineutrino experiments measure an antineutrino spectrum a few percent lower than expected from theoretical predictions. In this work we study the potential of low energy threshold reactor experiments in the context of a light sterile neutrino signal. We discuss the perspectives of the recently detected coherent elastic neutrino-nucleus scattering in future reactor antineutrino experiments. We find that the expectations to improve the current constraints on the mixing with sterile neutrinos are promising. We also analyze the measurements of antineutrino scattering off electrons from short distance reactor experiments. In this case, the statistics is not competitive with inverse beta decay experiments, although future experiments might play a role when compare it with the Gallium anomaly.

  10. Fast breeder reactors: can we learn from experience

    International Nuclear Information System (INIS)

    Keck, O.

    1981-01-01

    An economic analysis of FBRs, in particular the long-term benefits to be expected, with reference to the experience of the West German fast breeder reactor programme suggests ways of bringing more realism into governmental decisions on the development of new reactor types. It is suggested that if reactor manufacturers and utilities financed commercial-size demonstration plants from their own funds, then the government would get more realistic advice. (U.K.)

  11. Gas reactor and associated nuclear experience in the UK relevant to high temperature reactor engineering

    International Nuclear Information System (INIS)

    Beech, D.J.; May, R.

    2000-01-01

    In the UK, the NNC played a leading role in the design and build of all of the UK's commercial magnox reactors and advanced gas-cooled reactors (AGRs). It was also involved in the DRAGON project and was responsible for producing designs for large scale HTRs and other gas reactor designs employing helium and carbon dioxide coolants. This paper addresses the gas reactor experience and its relevance to the current HTR designs under development which use helium as the coolant, through the consideration of a representative sample of the issues addressed in the UK by the NNC in support of the AGR and other reactor programmes. Modern HTR designs provide unique engineering challenges. The success of the AGR design, reflected in the extended lifetimes agreed upon by the licensing authorities at many stations, indicates that these challenges can be successfully overcome. The UK experience is unique and provides substantial support to future gas reactor and high temperature engineering studies. (authors)

  12. Brookhaven Reactor Experiment Control Facility, a distributed function computer network

    International Nuclear Information System (INIS)

    Dimmler, D.G.; Greenlaw, N.; Kelley, M.A.; Potter, D.W.; Rankowitz, S.; Stubblefield, F.W.

    1975-11-01

    A computer network for real-time data acquisition, monitoring and control of a series of experiments at the Brookhaven High Flux Beam Reactor has been developed and has been set into routine operation. This reactor experiment control facility presently services nine neutron spectrometers and one x-ray diffractometer. Several additional experiment connections are in progress. The architecture of the facility is based on a distributed function network concept. A statement of implementation and results is presented

  13. Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    1978-09-01

    The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors

  14. Flux distribution measurements in the Bruce A unit 1 reactor

    International Nuclear Information System (INIS)

    Okazaki, A.; Kettner, D.A.; Mohindra, V.K.

    1977-07-01

    Flux distribution measurements were made by copper wire activation during low power commissioning of the unit 1 reactor of the Bruce A generating station. The distribution was measured along one diameter near the axial and horizontal midplanes of the reactor core. The activity distribution along the copper wire was measured by wire scanners with NaI detectors. The experiments were made for five configurations of reactivity control mechanisms. (author)

  15. Experience in Reviewing Small Modular Reactor Technology

    International Nuclear Information System (INIS)

    Ahmad Nabil Abdul Rahim; Alfred, S.L.; Phongsakorn, P.

    2015-01-01

    Malaysia is in the stage of conducting Preliminary Technical Feasibility Study for the Deployment of Small Modular Reactor (SMR). There are different types of SMR, some already under construction in Argentina (CAREM) and China (HTR-PM) - (light water reactor and high temperature reactor technologies), others with near-term deployment such as SMART in South Korea, ACP100 in China, mPower and NuScale in the US, and others with longer term deployment prospects (liquid-metal cooled reactor technologies). The study was mainly to get an overview of the technology available in the market. The SMR ranking in the study was done through listing out the most deployable technology in the market according to their types. As a new comer country, the proven technology with an excellent operation history will usually be the main consideration points. (author)

  16. Safety operation of training reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.

    2001-01-01

    There are three nuclear research reactors in the Czech Republic in operation now: light water reactor LVR-15, maximum reactor power 10 MW t , owner and operator Nuclear Research Institute Rez; light water zero power reactor LR-0, maximum reactor power 5 kW t , owner and operator Nuclear Research Institute Rez and training reactor VR-1 Sparrow, maximum reactor power 5 kW t , owner and operate Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague. The training reactor VR-1 Vrabec 'Sparrow', operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly it is designed for training the students of Czech universities, preparing the experts for the Czech nuclear programme, as well as for certain research work, and for information programmes in the nuclear programme, as well as for certain research work, and for information programmes in sphere of using the nuclear energy (public relations). (author)

  17. Review of irradiation experiments for water reactor safety research

    International Nuclear Information System (INIS)

    Tobioka, Toshiaki

    1977-02-01

    A review is made of irradiation experiments for water reactor safety research under way in both commercial power plants and test reactors. Such experiments are grouped in two; first, LWR fuel performance under normal and abnormal operating conditions, and second, irradiation effects on fracture toughness in LWR vessels. In the former are fuel densification, swelling, and the influence of power ramp and cycling on fuel rod, and also fuel rod behavior under accident conditions in in-reactor experiment. In the latter are the effects of neutron exposure level on the ferritic steel of pressure vessels, etc.. (auth.)

  18. The RERTR demonstration experiments program at the Ford Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wehe, D K; King, J S [Department of Nuclear Engineering, University of Michigan (United States)

    1983-08-01

    The purpose of this paper is to highlight a major part of the experimental work which is being carried out at the Ford Nuclear Reactor (FNR) in conjunction with the RERTR program. A demonstration experiments program has been developed to: 1) characterize the FNR in sufficient detail to discern and quantify neutronic differences between the high and low enriched cores; 2) provide the theoretical group with measurements to benchmark their calculations. As with any experimental program associated with a reactor, stringent constraints limit the experiments which can be performed. Some experiments are performed routinely on the FNR (such as control rod calibrations), and much data is already available. Unfortunately, the accuracy we demand precludes using much of this earlier data. And in many cases, the requirement of precise (and copious) data has led to either developing new techniques (as in the case of rhodium mapping and neutron diffraction) or to further refinements on existing methods (as in the case of spectral unfolding). Nevertheless, we have tried to stay within the realm of recognized, well-established experimental methods in order to assuage any doubts about measured differences between HEU and LEU core parameters. This paper describes the principal results of the experiments performed so far.

  19. Vibration behavior of PWR reactor internals Model experiments and analysis

    International Nuclear Information System (INIS)

    Assedo, R.; Dubourg, M.; Livolant, M.; Epstein, A.

    1975-01-01

    In the late 1971, the CEA and FRAMATOME decided to undertake a comprehensive joint program of studying the vibration behavior of PWR internals of the 900 MWe, 50 cycle, 3 loop reactor series being built by FRAMATOME in France. The PWR reactor internals are submitted to several sources of excitation during normal operation. Two main sources of excitation may effect the internals behavior: the large flow turbulences which could generate various instabilities such as: vortex shedding: the pump pressure fluctuations which could generate acoustic noise in the circuit at frequencies corresponding to shaft speed frequencies or blade passing frequencies, and their respective harmonics. The flow induced vibrations are of complex nature and the approach selected, for this comprehensive program, is semi-empirical and based on both theoretical analysis and experiments on a reduced scale model and full scale internals. The experimental support of this program consists of: the SAFRAN test loop which consists of an hydroelastic similitude of a 1/8 scale model of a PWR; harmonic vibration tests in air performed on full scale reactor internals in the manufacturing shop; the GENNEVILLIERS facilities which is a full flow test facility of primary pump; the measurements carried out during start up on the Tihange reactor. This program will be completed in April 1975. The results of this program, the originality of which consists of studying separately the effects of random excitations and acoustic noises, on the internals behavior, and by establishing a comparison between experiments and analysis, will bring a major contribution for explaining the complex vibration phenomena occurring in a PWR

  20. Reactor Engineering Department annual report (April 1, 1990 - March 31, 1991)

    International Nuclear Information System (INIS)

    1991-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1990 (April 1, 1990 - March 31, 1991). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  1. Reactor Engineering Department annual report (April 1, 1991-March 31, 1992)

    International Nuclear Information System (INIS)

    1992-08-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1991 (April 1, 1991-March 31, 1992). The major Department's programs promoted in the year are assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researchers on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative work to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  2. Reactor Engineering Department annual report (April 1, 1987 - March 31, 1988)

    International Nuclear Information System (INIS)

    1988-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1987 (April 1, 1987 - March 31, 1988). The major activities in the Department concerns the programs of the high temperature gas-cooled reactor, the high conversion light water reactor, the advanced fission reactor system and the fusion reactor at JAERI and the fast breeder reactor at PNC. The report contains the latest progress in nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control/diagnosis and robotics, as well as the new topics from this fiscal year on advanced reactors system design studies and technique developments related the facilities in the Department. Also described are the activities of the Research Committee on Reactor Physics. (author)

  3. Reactor Engineering Department annual report (April 1, 1988 - March 31, 1989)

    International Nuclear Information System (INIS)

    1989-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1988 (April 1, 1988 - March 31, 1989). The Department has promoted cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and also to PNC's fast reactor project. Other major Department's programs are the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Application of a high energy accelerator to the nuclear engineering is also preliminarily assessed. The report also contains the latest progress in various basic researches as nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/ diagnosis and technical developments related to the reactor physics facilities. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  4. Extensive utilization of training reactor VR-1

    International Nuclear Information System (INIS)

    Karel, Matejka; Lubomir, Sklenka

    2005-01-01

    This paper describes one of the main purposes of the VR-1 training reactor utilisation - i.e. extensive educational programme. The educational programme is intended for the training of university students (all technical universities in Czech Republic) and selected nuclear power plant personnel. At the present, students can go through more than 20 different experimental exercises. An attractive programme including demonstration of reactor operation is prepared also for high school students. Moreover, research and development works and information programmes proceed at the VR-1 reactor as well

  5. Current status of the Thai Research Reactor (TRR-1/M1)

    International Nuclear Information System (INIS)

    Chueinta, Siripone; Julanan, Mongkol; Charncanchee, Decharchai

    2006-01-01

    The first Thai Research Reactor, TRR-1 went critical on 27 October 1962 at the maximum power of 1 MW. It was located at Office of Atoms for Peace (OAP) in Bangkok. Since then, TRR-1 was continuously operated and eventually shut down in 1975. Plate type, high-enriched uranium (HEU) and U 3 O 8 A1 cladding were used as the reactor fuel. Light water was used as moderator and coolant as well. In 1975, because of the problem from fuel supplier and also to supporting the Treaty of Non Proliferation of Nuclear Weapon or NPT, TRR-1 was shut down for modification. The reactor core and control system were disassembled and replaced by TRIGA Mark III. A new core was a hexagonal core shape designed by General Atomics (GA). Afterwards, TRR-1 was officially renamed to the Thai Research Reactor-1/Modification 1 (TRR-1/M1). TRR-1/M1 is a multipurpose swimming pool type reactor with nominal power of 2 MW. The TRR-1/M1 uses uranium enriched at 20% in U-235 (LEU) and ZrH alloy as fuel. Light water is also used as coolant and moderator. At present, the reactor is operating with core No.14. The reactor has been serving for various kinds of utilization namely, radioisotope production, neutron activation analysis, beam experiments and reactor physics experiments. (author)

  6. Educational reactor-physics experiments with the critical assemble TCA

    Energy Technology Data Exchange (ETDEWEB)

    Tsutsui, Hiroaki; Okubo, Masaaki; Igashira, Masayuki [Tokyo Inst. of Tech. (Japan); Horiki, Oichiro; Suzaki, Takenori

    1997-10-01

    The Tank-Type Critical Assembly (TCA) of Japan Atomic Energy Research Institute is research equipment for light water reactor physics. In the present report, the lectures given to the graduate students of Tokyo Institute of Technology who participated in the educational experiment course held on 26-30 August at TCA are rearranged to provide useful information for those who will implement educational basic experiments with TCA in the future. This report describes the principles, procedures, and data analyses for (1) Critical approach and Exponential experiment, (2) Measurement of neutron flux distribution, (3) Measurement of power distribution, (4) Measurement of fuel rod worth distribution, and (5) Measurement of safety plate worth by the rod drop method. (author)

  7. Educational reactor-physics experiments with the critical assembly TCA

    International Nuclear Information System (INIS)

    Tsutsui, Hiroaki; Okubo, Masaaki; Igashira, Masayuki; Horiki, Oichiro; Suzaki, Takenori.

    1997-10-01

    The Tank-Type Critical Assembly (TCA) of Japan Atomic Energy Research Institute is research equipment for light water reactor physics. In the present report, the lectures given to the graduate students of Tokyo Institute of Technology who participated in the educational experiment course held on 26-30 August at TCA are rearranged to provide useful information for those who will implement educational basic experiments with TCA in the future. This report describes the principles, procedures, and data analyses for 1) Critical approach and Exponential experiment, 2) Measurement of neutron flux distribution, 3) Measurement of power distribution, 4) Measurement of fuel rod worth distribution, and 5) Measurement of safety plate worth by the rod drop method. (author)

  8. Occupational analysis for the Angra-1 reactor

    International Nuclear Information System (INIS)

    Moraes, A.

    1991-01-01

    Due to several modifications which were imposed to its time schedule during construction, the Angra-1 reactor did not enter to the grid in 1982 as it was initially foreseen. These modifications occurred due to an unforeseen scenario that was verified in steam generators (serie D-3, Westinghouse) of power stations with similar configurations which had been installed in other countries such as Ringhals-3 (Sweden), Almaraz-1 (Spain) and McGuine-1 (USA). So, among the main events that occurred in the Angra-1 reactor, which were of interest from the point of view of radiation protection, it could be pointed out the personnel monitoring, and the occupational exposure measurements at different reactor power, during the reactor fueling and during modification and tests performed at the steam generators and at ducts of the primary coolant circuit. (author)

  9. French experience in research reactor fuel transportation

    International Nuclear Information System (INIS)

    Raisonnier, Daniele

    1996-01-01

    Since 1963 Transnucleaire has safely performed a large number of national and international transports of radioactive material. Transnucleaire has also designed and supplied suitable packaging for all types of nuclear fuel cycle radioactive material from front-end and back-end products and for power or for research reactors. Transportation of spent fuel from power reactors are made on a regular and industrial basis, but this is not yet the case for the transport of spent fuel coming from research reactors. Each shipment is a permanent challenge and requires a reactive organization dealing with all the transportation issues. This presentation will explain the choices made by Transnucleaire and its associates to provide and optimize the corresponding services while remaining in full compliance with the applicable regulations and customer requirements. (author)

  10. PAHR experiments in the MELUSINE reactor

    International Nuclear Information System (INIS)

    Rousseau, D.; Dereymez, P.; Guyon, H.; Junod, E.; Ploujoux, M.; Tournebize, F.; Backs, H.

    1983-01-01

    After a hypothetical accident in a fast neutron reactor core, the nuclear fuel and construction materials melt partially. In several out-of-pile devices, the melting materials and the sodium coolant come to interact thermodynamically. In short, a few seconds after the accident a bed of debris immersed in sodium is formed on a plane of steel. The PAHR program has as principal objective to study the thermodynamic behaviour of this bed in the MELUSINE reactor, taking into account the most crucial parameters that rule the phenomena. More particularly, the aim is to draw attention to the bed behaviour beyond the fusion point of the steel up to the partial fusion of the fuel. The authors describe the CELIA capsule and its instrumentation; the operation conditions of the reactor and the coupling factor; the out-of-pile materials and their operation conditions. (Auth.)

  11. System Requirements Document for the Molten Salt Reactor Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Aigner, R.D.

    2000-04-01

    The purpose of the conversion process is to convert the {sup 233}U fluoride compounds that are being extracted from the Molten Salt Reactor Experiment (MSRE) equipment to a stable oxide for long-term storage at Bldg. 3019.

  12. From USA operation experience of industrial uranium-graphite reactors

    International Nuclear Information System (INIS)

    Burdakov, N.S.

    1996-01-01

    The review on materials, presented by a group of the USA specialists at the seminar in Moscow on October 9-11, 1995 is considered. The above specialists shared their experience in operation of the Hanford industrial reactors, aimed at plutonium production for atomic bombs. The purpose of the above visit consisted in providing assistance to the Russian specialists by evaluation and modernization of operational conditions safety improvement of the RBMK type reactors. Special attention is paid to the behaviour of the graphite lining and channel tubes with an account of possible channel power interaction with the reactor structural units. The information on the experience of the Hanford reactor operation may be useful for specialists, operating the RBMK type reactors

  13. Substantiation of physical concepts of fast reactors in Russia: experience and prospects

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, P.N. [Russian Research Center ' Kurchatov Institute' (RRC KI), 1, Kurchatov Sq., Moscow, 123182 (Russian Federation); Vasiliev, B.A. [Experimental Design Bureau of Machine Building (OKBM) 15, Burnakovskiy Pr., N. Novgorod, 603074 (Russian Federation); Kormilitsyn, M.V. [State Scientific Center of Russian Federation - Research Institute of Atomic Reactors (NIIAR) Dimitrovgrad-10, Ulianovsk Reg., 433510 (Russian Federation); Lopatkin, A.V. [N.A. Dollezhal Research and Development Institute of Power Engineering (NIKIET) 2/8, M. Krasnoselskaya Str., Moscow, 107140 (Russian Federation); Seleznev, E.F. [All-Russian Research Institute for Nuclear Power Plant Operation (VNIIAES) 25, Ferganskaya, Moscow, 109507 (Russian Federation); Khomyakov, Yu.S.; Tsybulia, A.M. [State Scientific Center of the Russian Federation - A. I. Leypunsky Institute for Physics and Power Engineering (SSC RF- IPPE) 1, Bondarenko Sq., Obninsk, Kaluga Reg., 249033 (Russian Federation); Tocheny, L.V. [International Science and Technology Center (ISTC) 32-34 Krasnoproletarskaya Ulitsa, Moscow, 127473 (Russian Federation)

    2008-07-01

    The fast reactor concept in Russia has accumulated unique experience, since its advent in the 1950's and up to the present, from the creation of the first experimental installation BR-1, experimental reactors BR-5 and BOR-60, the pilot industrial reactors BN-350 in Kazakhstan and up to the BN-600 at Beloyarsk Atomic Power Station. Investigations on the first experimental installations BR-1 and BR-5/-10 proved the propriety of the idea that it is possible to create nuclear reactors that can produce more nuclear fuel than they consume, i.e. the idea of breeding. The architecture of such reactors was also designed, producing a current leader among fast reactors with sodium coolant and oxide uranium-plutonium fuel. Operational experience of BOR-60, BN-350 and, particularly, BN-600 confirmed the engineering and technical feasibility of the concept of fast reactors, the possibility for its realization both for power production and for certain other purposes as well, such as desalinisation of sea water (BN-350) and for radionuclide production (BN-350, BN-600), and it enabled the development and verification of different models, computer methods and codes. The paper presents a review of experience in the creation of plants with fast reactors, scientific research on these installations, principal results, the current status of experimental data analysis, and prospective directions in the development of fast reactors and the corresponding experimental basis in Russia. (authors)

  14. Annual report on JEN-1 reactor; Informe periodico del Reactor JEN-1 correspondiente al ano 1971

    Energy Technology Data Exchange (ETDEWEB)

    Montes, J

    1972-07-01

    In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

  15. Safety Technology Research Program in the field of pressurized water reactors. 1. Technical report on advancement project RS 36/2. Emergency cooling program service life experiments: reflooding experiments involving the primary loop systems

    International Nuclear Information System (INIS)

    Schweickert, H.; Kremin, H.; Mandl, R.; Riedle, V.; Ruthrof, K.; Sarkar, J.; Schmidt, H.

    The reflooding of the hot reactor core is to be examined for a pressurized water reactor (PWR), using a model of the entire primary loop system. The scale of the model is to be 1:340 in cross-section, with the heights represented full-scale. In addition to the goals of the project, a description of the test facility, including data collection and control equipment is presented. The instrumentation, the planned test program and the test procedure are briefly set forth

  16. DOE's foreign research reactor transportation services contract: Perspective and experience

    International Nuclear Information System (INIS)

    Patterson, John

    1997-01-01

    DOE committed to low- and moderate-income countries participating in the foreign research reactor spent fuel returns program that the United States government would provide for the transportation of the spent fuel. In fulfillment of that commitment, DOE entered into transportation services contracts with qualified, private-sector firms. NAC will discuss its experience as a transportation services provider, including range of services available to the foreign reactors, advantages to DOE and to the foreign research reactors, access to contract services by high income countries and potential advantages, and experience with initial tasks performed under the contract. (author)

  17. Tritium experience in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A.; Hogan, J.

    1998-01-01

    Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors

  18. Operation experience at the Neuherberg Research Reactor (FRN) with several modifications of reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Demmeler, M; Rau, G [Gesellschaft fuer Strahlen- und Umweltforschung mbH, Neuherberg (Germany)

    1974-07-01

    Since the first full power operation in September 1972 up till now (Dec. 1973) the TRIGA Mark III reactor FRN has run more than 500 MWh in steady state operation and has been pulsed for 265 times. During startup experiments, neutron- and gamma-flux mapping has been performed with special technical devices in the core and in several irradiation positions, mainly in the thermal column and in the exposure room. Furthermore reactivity values of each fuel element have been measured at full power of 1 MW, thus enabling a more accurate burnup calculation. Troubles with the rotary specimen rack occurred at power rates above 280 kW; here, the lazy susan stuck, caused by thermal stress. Thus it will be replaced by a hydraulic-operated type, which has been developed at the TRIGA reactor Heidelberg. In order to increase irradiation capacity, a new core configuration has been set up a few months ago, replacing several fuel-reflector-elements by irradiation tubes within the grid-plate positions E-22, G-2, G-17 and G-36. Four additional fuel elements had to be inserted to compensate for the resulting reactivity losses. The original plan of regaining sufficient excess-reactivity by inserting a fuel element in grid-plate position A-l failed because of local boiling in the center of the core by 1 MW-operation. Experiments at the reactor started with the begin of routine-operation in September 1973. Up till now, a total of 450 neutron- and gamma- irradiations have been performed, mainly for neutron-activations. (author)

  19. A plan of reactor physics experiments for reduced-moderation water reactors with MOX fuel in TCA

    International Nuclear Information System (INIS)

    Shimada, Shoichiro; Akie, Hiroshi; Suzaki, Takenori; Okubo, Tutomu; Usui, Shuji; Shirakawa, Toshihisa; Iwamura, Takamiti; Kugo, Teruhiko; Ishikawa, Nobuyuki

    2000-06-01

    The Reduced-Moderation Water Reactor (RMWR) is one of the next generation water-cooled reactors which aim at effective utilization of uranium resource, high burn-up, long operation cycle, and plutonium multi-recycle. For verification of the feasibility, negative void reactivity coefficient and conversion ratio more than 1.0 must be confirmed. Critical Experiments performed so far in Eualope and Japan were reviewed, and no useful data are available for RMWR development. Critical experiments using TCA (Tank Type Critical Assembly) in JAERI are planned. MOX fuel rods should be prepared for the experiments and some modifications of the equipment are needed for use of MOX fuel rods. This report describes the preliminary plan of physics experiments. The number of MOX fuel rods used in the experiments are obtained by calculations and the modification of the equipment for the experiments are shown. (author)

  20. R+D works for the further development of high temperature reactors. (1) Captive bearing experiments for active magnetic bearings. (2) Captive bearing test for HTR blowers

    International Nuclear Information System (INIS)

    1991-01-01

    When using active magnetic bearings as blower shaft bearings, blower motors and bearings must be protected against mechanical damage in case of faults (example: total electrical supply failure due to the supply cables breaking). So-called captive bearings are provided, in order to be able to shut the blowers down safely in such faults. These captive bearings are roller bearings which are additionally fitted in the area of the blower shaft bearings, to prevent mechanical contact between the blower rotor and stator. As there was little experience available for the given boundary conditions, such as - speed, - acceleration, - bearing load, - bearing dimensions, - ambient conditions, appropriate development and tests had to be carried out. It was important to determine suitable captive bearings and the necessary ambient conditions, which will make it possible to support the failures of the magnetic bearings to be expected in 40 years' operation of the reactor without damage and to meet the requirements of the captive bearings. (orig./GL) [de

  1. New human machine interface for VR-1 training reactor

    International Nuclear Information System (INIS)

    Kropik, M.; Matejka, K.; Sklenka, L.; Chab, V.

    2002-01-01

    The contribution describes a new human machine interface that was installed at the VR-1 training reactor. The human machine interface update was completed in the summer 2001. The human machine interface enables to operate the training reactor. The interface was designed with respect to functional, ergonomic and aesthetic requirements. The interface is based on a personal computer equipped with two displays. One display enables alphanumeric communication between a reactor operator and the control and safety system of the nuclear reactor. Messages appear from the control system, the operator can write commands and send them there. The second display is a graphical one. It is possible to represent there the status of the reactor, principle parameters (as power, period), control rods' positions, the course of the reactor power. Furthermore, it is possible to set parameters, to show the active core configuration, to perform reactivity calculations, etc. The software for the new human machine interface was produced in the InTouch developing environment of the WonderWare Company. It is possible to switch the language of the interface between Czech and English because of many foreign students and visitors at the reactor. The former operator's desk was completely removed and superseded with a new one. Besides of the computer and the two displays, there are control buttons, indicators and individual numerical displays of instrumentation there. Utilised components guarantee high quality of the new equipment. Microcomputer based communication units with proper software were developed to connect the contemporary control and safety system with the personal computer of the human machine interface and the individual displays. New human machine interface at the VR-1 training reactor improves the safety and comfort of the reactor utilisation, facilitates experiments and training, and provides better support of foreign visitors.(author)

  2. Description of the french graphite reactor and of the experiments performed in 1956

    International Nuclear Information System (INIS)

    Bussac, J.; Leduc, C.; Zaleski, C.P.

    1957-01-01

    This paper is an introduction to the experiments performed on the G1 reactor, experiments fully described in the papers following (670 'B to P'). The main results are given together with some comments. The neutronic parameters of the core, a description of the most important structures, and a few words of the tests leading to normal operation of the reactor under load complete our survey. (author) [fr

  3. Operating experience feedback from safety significant events at research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shokr, A.M. [Atomic Energy Authority, Abouzabal (Egypt). Egypt Second Research Reactor; Rao, D. [Bhabha Atomic Research Centre, Mumbai (India)

    2015-05-15

    Operating experience feedback is an effective mechanism to provide lessons learned from the events and the associated corrective actions to prevent recurrence of events, resulting in improving safety in the nuclear installations. This paper analyzes the events of safety significance that have been occurred at research reactors and discusses the root causes and lessons learned from these events. Insights from literature on events at research reactors and feedback from events at nuclear power plants that are relevant to research reactors are also presented along with discussions. The results of the analysis showed the importance of communication of safety information and exchange of operating experience are vital to prevent reoccurrences of events. The analysis showed also the need for continued attention to human factors and training of operating personnel, and the need for establishing systematic ageing management programmes of reactor facilities, and programmes for safety management of handling of nuclear fuel, core components, and experimental devices.

  4. Experiences in stability testing of boiling water reactors

    International Nuclear Information System (INIS)

    March-Leuba, J.; Otaduy, P.J.

    1986-01-01

    The purpose of this paper is to summarize experiences with boiling water reactor (BWR) stability testing using noise analysis techniques. These techniques have been studied over an extended period of time, but it has been only recently that they have been well established and generally accepted. This paper contains first a review of the problem of BWR neutronic stability, focusing on its physical causes and its effects on reactor operation. The paper also describes the main techniques used to quantify, from noise measurements, the reactor's stability in terms of a decay ratio. Finally, the main results and experiences obtained from the stability tests performed at the Dresden and the Browns Ferry reactors using noise analysis techniques are summarized

  5. Operating experience feedback from safety significant events at research reactors

    International Nuclear Information System (INIS)

    Shokr, A.M.

    2015-01-01

    Operating experience feedback is an effective mechanism to provide lessons learned from the events and the associated corrective actions to prevent recurrence of events, resulting in improving safety in the nuclear installations. This paper analyzes the events of safety significance that have been occurred at research reactors and discusses the root causes and lessons learned from these events. Insights from literature on events at research reactors and feedback from events at nuclear power plants that are relevant to research reactors are also presented along with discussions. The results of the analysis showed the importance of communication of safety information and exchange of operating experience are vital to prevent reoccurrences of events. The analysis showed also the need for continued attention to human factors and training of operating personnel, and the need for establishing systematic ageing management programmes of reactor facilities, and programmes for safety management of handling of nuclear fuel, core components, and experimental devices.

  6. Experiment prediction for Loft Nonnuclear Experiment L1-4

    International Nuclear Information System (INIS)

    White, J.R.; Berta, V.T.; Holmstrom, H.L.O.

    1977-04-01

    A computer analysis, using the WHAM and RELAP4 computer codes, was performed to predict the LOFT system thermal-hydraulic response for Experiment L1-4 of the nonnuclear (isothermal) test series. Experiment L1-4 will simulate a 200 percent double-ended offset shear in the cold leg of a four-loop large pressurized water reactor. A core simulator will be used to provide a reactor vessel pressure drop representative of the LOFT nuclear core. Experiment L1-4 will be initiated with a nominal isothermal primary coolant temperature of 282.2 0 C, a pressurizer pressure of 15.51 MPa, and a primary coolant flow of 270.9 kg/s. In general, the predictions of saturated blowdown for Experiment Ll-4 are consistent with the expected system behavior, and predicted trends agree with results from Semiscale Test S-01-4A, which simulated the Ll-4 experiment conditions

  7. Operating experiences at the Finnish TRIGA reactor

    International Nuclear Information System (INIS)

    Salmenhaara, Seppo

    1988-01-01

    The Finnish TRIGA reactor has been in operation since March 1962. There are still 57 original Al-clad fuel elements in the core. So far we have had only two fuel cladding failures in 1981 and 1988. The first one was an Al-clad element and the second one a SS-clad. The low rate of fuel cladding failures has made it possible to use continuously also the Al-clad fuel elements. Although some conventional irradiations of certain type have been repeated successfully tens of times, new and unexpected incidents can still take place. As an example an event of a leaking irradiation capsule is described

  8. The primary circuit of the dragon high temperature reactor experiment

    International Nuclear Information System (INIS)

    Simon, R.

    2005-01-01

    The 20 MWth Dragon Reactor Experiment was the first HTGR (High Temperature Gas-cooled Reactor) with coated particle fuel. Its purpose was to test fuel and materials for the High Temperature Reactor programmes pursued in Europe 40 years ago. This paper describes the design and construction of the primary (helium) circuit. It summarizes the main design objectives, lists the performance data and explains the flow paths of the heat removal and helium purification systems. The principal circuit accidents postulated are discussed and the choice of the main construction materials is given. (author)

  9. Power cycling experiments in INR-TRIGA-SSR Reactor

    International Nuclear Information System (INIS)

    Dumitru, M.

    2008-01-01

    The in-reactor experimental program started this summer with some power cycling experiments to provide date on fuel behaviour under abnormal reactor operating conditions. The paper describes the irradiation device, its operational features and an original 'under-flux' movement system. Also, there are presented main data of irradiation device (pressure, flow, temperature, construction), in-pile section, location, sample, instrumentation, experimental sequences and operating data of Interest for the experimenters. (author)

  10. Decommissioning the Los Alamos Molten Plutonium Reactor Experiment (LAMPRE I)

    International Nuclear Information System (INIS)

    Harper, J.R.; Garde, R.

    1981-11-01

    The Los Alamos Molten Plutonium Reactor Experiment (LAMPRE I) was decommissioned at the Los Alamos National Laboratory, Los Alamos, New Mexico, in 1980. The LAMPRE I was a sodium-cooled reactor built to develop plutonium fuels for fast breeder applications. It was retired in the mid-1960s. This report describes the decommissioning procedures, the health physics programs, the waste management, and the costs for the operation

  11. Performance improvement of the Annular Core Pulse Reactor for reactor safety experiments

    International Nuclear Information System (INIS)

    Reuscher, J.A.; Pickard, P.S.

    1976-01-01

    The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past several years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. The new features of the improvements are described

  12. Operation and maintenance experience at the General Atomic Company's TRIGA reactor facility at San Diego, California

    International Nuclear Information System (INIS)

    Whittemore, W.L.; Stout, W.A.; Shoptaugh, J.R.; Chesworth, R.H.

    1982-01-01

    Since the startup of the original 250 kW TRIGA Mark I reactor in 1958, General Atomic Company has accumulated nearly 24 years of operation and maintenance experience with this type of reactor. In addition to the nearly 24 years of experience gained on the Mark I, GA has operated the 1.5 MW Advanced Prototype Test Reactor (Mark F) for 22 years and operated a 2 MW below-ground TRIGA Mark III for five years. Information obtained from normal and abnormal operation are presented. (author)

  13. Thermal hydraulic reactor safety analyses and experiments

    International Nuclear Information System (INIS)

    Holmstroem, H.; Eerikaeinen, L.; Kervinen, T.; Kilpi, K.; Mattila, L.; Miettinen, J.; Yrjoelae, V.

    1989-04-01

    The report introduces the results of the thermal hydraulic reactor safety research performed in the Nuclear Engineering Laboratory of the Technical Research Centre of Finland (VTT) during the years 1972-1987. Also practical applications i.e. analyses for the safety authorities and power companies are presented. The emphasis is on description of the state-of-the-art know how. The report describes VTT's most important computer codes, both those of foreign origin and those developed at VTT, and their assessment work, VTT's own experimental research, as well as international experimental projects and other forms of cooperation VTT has participated in. Appendix 8 contains a comprehensive list of the most important publications and technical reports produced. They present the content and results of the research in detail.(orig.)

  14. COUNTERCURRENT FLOW LIMITATION EXPERIMENTS AND MODELING FOR IMPROVED REACTOR SAFETY

    International Nuclear Information System (INIS)

    Vierow, Karen

    2008-01-01

    This project is investigating countercurrent flow and 'flooding' phenomena in light water reactor systems to improve reactor safety of current and future reactors. To better understand the occurrence of flooding in the surge line geometry of a PWR, two experimental programs were performed. In the first, a test facility with an acrylic test section provided visual data on flooding for air-water systems in large diameter tubes. This test section also allowed for development of techniques to form an annular liquid film along the inner surface of the 'surge line' and other techniques which would be difficult to verify in an opaque test section. Based on experiences in the air-water testing and the improved understanding of flooding phenomena, two series of tests were conducted in a large-diameter, stainless steel test section. Air-water test results and steam-water test results were directly compared to note the effect of condensation. Results indicate that, as for smaller diameter tubes, the flooding phenomena is predominantly driven by the hydrodynamics. Tests with the test sections inclined were attempted but the annular film was easily disrupted. A theoretical model for steam venting from inclined tubes is proposed herein and validated against air-water data. Empirical correlations were proposed for air-water and steam-water data. Methods for developing analytical models of the air-water and steam-water systems are discussed, as is the applicability of the current data to the surge line conditions. This report documents the project results from July 1, 2005 through June 30, 2008

  15. Licensing experience of the HTR-10 test reactor

    International Nuclear Information System (INIS)

    Sun, Y.; Xu, Y.

    1996-01-01

    A 10MW high temperature gas-cooled test reactor (HTR-10) is now being projected by the Institute of Nuclear Energy Technology within China's National High Technology Programme. The Construction Permit of HTR-10 was issued by the Chinese nuclear licensing authority around the end of 1994 after a period of about one year of safety review of the reactor design. HTR-10 is the first high temperature gas-cooled reactor (HTGR) to be constructed in China. The purpose of this test reactor project is to test and demonstrate the technology and safety features of the advanced modular high temperature reactor design. The reactor uses spherical fuel elements with coated fuel particles. The reactor unit and the steam generator unit are arranged in a ''side-by-side'' way. Maximum fuel temperature under the accident condition of a complete loss of coolant is limited to values much lower than the safety limit set for the fuel element. Since the philosophy of the technical and safety design of HTR-10 comes from the high temperature modular reactor design, the reactor is also called the Test Module. HTR-10 represents among others also a licensing challenge. On the one side, it is the first helium reactor in China, and there are less licensing experiences both for the regulator and for the designer. On the other side, the reactor design incorporates many advanced design features in the direction of passive or inherent safety, and it is presently a world-wide issue how to treat properly the passive or inherent safety design features in the licensing safety review. In this presentation, the licensing criteria of HTR-10 are discussed. The organization and activities of the safety review for the construction permit licensing are described. Some of the main safety issues in the licensing procedure are addressed. Among these are, for example, fuel element behaviour, source term, safety classification of systems and components, containment design. The licensing experiences of HTR-10 are of

  16. Liquid metal cooled reactors: Experience in design and operation

    International Nuclear Information System (INIS)

    2007-12-01

    on key fast reactor technology aspects in an integrative sense useful to engineers, scientists, managers, university students and professors. This publication has been prepared to contribute toward the IAEA activity to preserve the knowledge gained in the liquid metal cooled fast reactor (LMFR) technology development. This technology development and experience include aspects addressing not only experimental and demonstration reactors, but also all activities from reactor construction to decommissioning. This publication provides a survey of worldwide experience gained over the past five decades in LMFR development, design, operation and decommissioning, which has been accumulated through the IAEA programmes carried out within the framework of the TWG-FR and the Agency's INIS and NKMS

  17. Reactor engineering department annual report. April 1, 1993-March 31, 1994

    International Nuclear Information System (INIS)

    1994-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1993 (April 1, 1993-March 31, 1994). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees organized by the Department are also summarized in this report. (author)

  18. Oscillation experiments techniques in CEA Minerve experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Antony, M.; Di-Salvo, J.; Pepino, A.; Bosq, J. C.; Bernard, D.; Leconte, P.; Hudelot, J. P.; Lyoussi, A. [CEA CADARACHE, DEN/DER/SPEx, 13108 Saint Paul-lez-Durance (France)

    2009-07-01

    This paper deals with experiments in the Minerve pool Zero Power Reactor. Minerve is mainly devoted to neutronics studies, in view to improve the calculation routes by reducing the uncertainties of the experimental databases for nuclides arising in plutonium and wastes management. Minerve experimental measurement programs are performed by using the oscillation technique. This experimental technique consists in a periodic insertion and extraction of samples containing the nuclide of interest in a well characterized neutron spectrum. The reactivity variation of the sample is compensated by a calibrated rotary automatic pilot using cadmium sectors. The normal accuracy for measurements of small-worth samples in Minerve by using such a technique is about 3% for absolute reactivity worth, including the uncertainties on the material balance and on the calibration step. Reactivity effects of less than 1.5 cent can be measured. The OSMOSE and the OCEAN programs have been carried out since 2005 and will last until 2011. These programs aim at improving, in different neutron spectra, the absorption cross sections of respectively a majority of the separated heavy nuclides from {sup 232}Th to {sup 245}Cm appearing during the reactor and the fuel cycle physics, and of current and future types of absorbers as Gd, Hf, Er, Dy and Eu. (authors)

  19. Instrumentation renewal at the FIR 1 research reactor in Finland

    International Nuclear Information System (INIS)

    Bars, Bruno; Kall, Leif

    1982-01-01

    The Finnish TRIGA Mark II reactor (FIR 1 100 kW, later 250 kW steady state power and pulsing capability up to 250 MW) has been in operation for 20 years. The reactor is the only research reactor in Finland and is an important research training and service facility, which obviously will be operated for 10...20 years ahead. The mechanical parts of the reactor are in good shape. Some minor modifications have previously been made in the instrumentation. However, the original instrumentation could hardly have been used for 10...20 years ahead without extensive modifications and modernization. After a careful evaluation and planning process the whole reactor instrumentation was renewed in 1981 at a cost of about 400 000 dollar. The renewal was carried out in cooperation with the Central Research Institute for Physics (KFKI) at the Hungarian Academy of Sciences, which delivered the nuclear part of the instrumentation and with the Finnish company Valmet Oy Instrument Works, which delivered the conventional instrumentation, including the automatic power control system and the control console. The instrumentation, which is located in-a new isolated control room is based on modern industrial standard modular units with standardized signal ranges, electronic testing possibilities, galvanically isolated outputs etc. The instrument renewal project was brought successfully to completion in November 1981 after only about 10 working days of shut down time. The reactor is now in routine operation and the experiences gained from the new instrumentation are excellent. (author)

  20. Dry cooling tower operating experience in the LOFT reactor

    International Nuclear Information System (INIS)

    Hunter, J.A.

    1980-01-01

    A dry cooling tower has been uniquely utilized to dissipate heat generated in a small experimental pressurized water nuclear reactor. Operational experience revealed that dry cooling towers can be intermittently operated with minimal wind susceptibility and water hammer occurrences by cooling potential steam sources after a reactor scram, by isolating idle tubes from the external atmosphere, and by operating at relatively high pressures. Operating experience has also revealed that tube freezing can be minimized by incorporating the proper heating and heat loss prevention features

  1. Non-standard interaction effects at reactor neutrino experiments

    International Nuclear Information System (INIS)

    Ohlsson, Tommy; Zhang, He

    2009-01-01

    We study non-standard interactions (NSIs) at reactor neutrino experiments, and in particular, the mimicking effects on θ 13 . We present generic formulas for oscillation probabilities including NSIs from sources and detectors. Instructive mappings between the fundamental leptonic mixing parameters and the effective leptonic mixing parameters are established. In addition, NSI corrections to the mixing angles θ 13 and θ 12 are discussed in detailed. Finally, we show that, even for a vanishing θ 13 , an oscillation phenomenon may still be observed in future short baseline reactor neutrino experiments, such as Double Chooz and Daya Bay, due to the existences of NSIs

  2. Perspectives on reactor safety. Revision 1

    International Nuclear Information System (INIS)

    Haskin, F.E.; Hodge, S.A.

    1997-11-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course

  3. Perspectives on reactor safety. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A. [Oak Ridge National Lab., TN (United States). Engineering Technology Div.

    1997-11-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  4. Decommissioning and decontrolling the R1-reactor

    International Nuclear Information System (INIS)

    Bergman, C.; Holmberg, B.T.

    1985-01-01

    Sweden's first nuclear reactor - the research reactor R1 - situated in bedrock under the Royal Technical Institute of Stockholm, has in the period 1981-1983 been subject to a complete decommissioning. The National Institute for Radiation Protection has followed the work in detail, and has after the completion of the decommissioning performed measurements of radioactivity on site. The report gives an account of the work the Institute has done in preparation for- and during decommissioning and specifically report on the measurements for classification of the local as free for non-nuclear use. (aa)

  5. Radiological characterisation experience with Magnox reactors

    International Nuclear Information System (INIS)

    Westall, Bill A.; Towton, Barbara

    2012-01-01

    At the end of generation, power reactors will be decommissioned. Whether decommissioning is prompt or deferred, knowledge of the radioactive inventory of plant and structures is needed to develop and underpin the decommissioning strategy. As decommissioning progresses the level of detail required for the radioactive inventory increases as more specific and detailed questions need answering. Failure to adequately characterise will result in increased costs and project overruns due to missing optimal solutions, over pessimistic assumptions or unforeseen problems and regulatory issues. Radiological characterisation for decommissioning of Magnox power stations in the UK has been in progress for over a quarter of a century. Firstly measurements and calculations were carried out to develop a strategy. These have been followed by measurements to determine radioactive inventories of waste streams and packages or to allow decontamination of structures and most recently for partial de-licensing of sites. Some examples of the work carried out for the Magnox stations will be given, ranging from the neutron activation calculations to estimate the radioactive inventory within a bio-shield and measurements to validate them. Various plant and structures where the radioactive inventory is due to contamination have been characterised by measurements and examples for boilers and cooling ponds will be discussed. Various routine and ad-hoc measurements and shielding assessments have been performed on waste forms to help satisfy conditions for acceptance for disposal or exemption, which will be reviewed. Finally the measurements for de-licensing and the successful application of Data Quality Objectives will be addressed. (authors)

  6. Annual meeting on nuclear technology 1980. Technical meeting: Operating experiences

    International Nuclear Information System (INIS)

    1980-01-01

    In addition to general experiences, experiences in reactor operation with relation to the Phenix reactor, KNK-2 reactor, the AVR reactor, the BWR-type KKI-reactor, the Philippsburg-1 reactor and the Obrigheim reactor are described. (DG) [de

  7. Performance of a sphere-pac mixed carbide fuel pin irradiated in the Dounreay Fast Reactor (DFR 527/1 experiment)

    International Nuclear Information System (INIS)

    Bischoff, K.; Smith, L.; Stratton, R.W.

    1980-10-01

    The DFR 527/1 experiment was the first irradiation of EIR sphere-pac uranium-plutonium mixed carbide fuel in a fast flux. The experiment has been successfully irradiated to a burn-up of 7.3% FIMA at ratings between 45 and 62 kW m - 1 and clad temperatures between 300 and 600 0 C. Restructuring and elemental redistribution has been found to be similar to the pattern established for pellet type fuel and follows effects seen in earlier sphere-pac carbide tests. Gas release of 12-14% has been measured. A preliminary comparison of radial temperature distribution calculations using a first version of the fuel behaviour modelling code SPECKLE with the actual metallography has been attempted. (Auth.)

  8. Refurbishment of Pakistan research reactor (PARR-1) for stainless steel lining of the reactor pool

    International Nuclear Information System (INIS)

    Salahuddin, A.; Israr, M.; Hussain, M.

    2002-01-01

    Pakistan Research Reactor-1 (PARR-1) is a pool-type research reactor. Reactor aging has resulted in the increase of water seepage from the concrete walls of the reactor pool. To stop the seepage, it was decided to augment the existing pool walls with an inner lining of stainless steel. This could be achieved only if the pool walls could be accessed unhindered and without excessive radiation doses. For this purpose a partial decommissioning was done by removing all active core components including standard/control fuel elements, reflector elements, beam tubes, thermal shield, core support structure, grid plate and the pool's ceramic tiles, etc. An overall decommissioning program was devised which included procedures specific to each item. This led to the development of a fuel transport cask for transportation, and an interim fuel storage bay for temporary storage of fuel elements (until final disposal). The safety of workers and the environment was ensured by the use of specially designed remote handling tools, appropriate shielding and pre-planned exposure reduction procedures based on the ALARA principle. During the implementation of this program, liquid and solid wastes generated were legally disposed of. It is felt that the experience gained during the refurbishment of PARR-1 to install the stainless steel liner will prove useful and better planning and execution for the future decommissioning of PARR-1, in particular, and for other research reactors like PARR-2 (27 kW MNSR), in general. Furthermore, due to the worldwide activities on decommissioning, especially those communicated through the IAEA CRP on 'Decommissioning Techniques for Research Reactors', the importance of early planning has been well recognized. This has made possible the implementation of some early steps like better record keeping, rehiring of trained manpower, and creation of interim and final waste storage. (author)

  9. Investigating the spectral anomaly with different reactor antineutrino experiments

    Directory of Open Access Journals (Sweden)

    C. Buck

    2017-02-01

    Full Text Available The spectral shape of reactor antineutrinos measured in recent experiments shows anomalies in comparison to neutrino reference spectra. New precision measurements of the reactor neutrino spectra as well as more complete input in nuclear data bases are needed to resolve the observed discrepancies between models and experimental results. This article proposes the combination of experiments at reactors which are highly enriched in U235 with commercial reactors with typically lower enrichment to gain new insights into the origin of the anomalous neutrino spectrum. The presented method clarifies, if the spectral anomaly is either solely or not at all related to the predicted U235 spectrum. Considering the current improvements of the energy scale uncertainty of present-day experiments, a significance of three sigma and above can be reached. As an example, we discuss the option of a direct comparison of the measured shape in the currently running Double Chooz near detector and the upcoming Stereo experiment. A quantitative feasibility study emphasizes that a precise understanding of the energy scale systematics is a crucial prerequisite in recent and next generation experiments investigating the spectral anomaly.

  10. Tokamak fusion test reactor FELIX plate experiment

    International Nuclear Information System (INIS)

    Hua, T.O.; Nygren, R.E.; Turner, L.R.

    1986-01-01

    For a conducting material exposed to both a time-varying and a static magnetic field, such as a limiter blade in a tokamak, the induced eddy currents and the deflection arising from those eddy currents can be strongly coupled. The coupling effects reduce the currents and deflections markedly, sometimes an order of magnitude, from the values predicted if coupling is neglected. A series of experiments to study current-deflection coupling were performed using the Fusion Electromagnetic Inductance Experiment (FELIX) facility at Argonne National Laboratory. Magnetic damping and magnetic stiffness resulting from the coupling are discussed, and analytical expressions for induced eddy current and rigid body rotation in the FELIX plate experiment are compared with the experimental results. Predictions for the degree of coupling based on various parameters are made using the analytical model

  11. ORNL fusion reactor shielding integral experiments

    International Nuclear Information System (INIS)

    Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M.; Chapman, G.T.

    1980-01-01

    Integral experiments that measure the neutron and gamma-ray energy spectra resulting from the attenuation of approx. 14 MeV T(D,n) 4 He reaction neutrons in laminated slabs of stainless steel type 304, borated polyethylene, and a tungsten alloy (Hevimet) and from neutrons streaming through a 30-cm-diameter iron duct (L/D = 3) imbedded in a concrete shield have been performed. The facility, the NE-213 liquid scintillator detector system, and the experimental techniques used to obtain the measured data are described. The two-dimensional discrete ordinates radiation transport codes, calculational models, and nuclear data used in the analysis of the experiments are reviewed

  12. Liquid jet experiments: relevance to inertial confinement fusion reactors

    International Nuclear Information System (INIS)

    Hoffman, M.A.

    1981-01-01

    In order to try to find a reactor design which offered protection against neutron damage, studies were undertaken at LLNL (the Lawrence Livermore National Laboratory) of self-healing, renewable liquid-wall reactor concepts. In conjuction with these studies, were done a seris of small-scale aer jet experiments were done over the past several years at UCD (University of California, Davis Campus) to simulate the behavior of liquid lithium (or lithium-lead) jets in these liquid-wall fusion reactor concepts. Extropolating the results of these small-scale experiments to the large-scale lithium jets, tentatively concluded that the lithium jet can be re-established after the microexplosion, and with careful design the jets should not breakup due to instabilities during the relatively quiscent period between MICROEXPLOSIONS

  13. An Overview of the International Reactor Physics Experiment Evaluation Project

    International Nuclear Information System (INIS)

    Briggs, J. Blair; Gulliford, Jim

    2014-01-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties associated with advanced modeling and simulation accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Data provided by those two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades. An overview of the IRPhEP and a brief update of the ICSBEP are provided in this paper.

  14. Reactor dynamics experiment of nuclear ship Mutsu using pseudo random signal (II). The second experiment

    International Nuclear Information System (INIS)

    Hayashi, Koji; Shimazaki, Junya; Nabeshima, Kunihiko; Ochiai, Masaaki; Shinohara, Yoshikuni; Inoue, Kimihiko.

    1995-01-01

    In order to investigate dynamics of the reactor plant of the nuclear ship Mutsu, the second reactor noise experiment using pseudo random binary sequences (PRBS) was performed on August 30, 1991 in the third experimental navigation. The experiments using both reactivity and load disturbances were performed at 50% of reactor power and under a quiet sea condition. Each PRBS was applied by manual operation of the control rod or the main steam valve. Various signals of the plant responses and of the acceleration of ship motion were measured. Furthermore, natural reactor noise signals were measured after each PRBS experiment in order to evaluate the effects of the PRBS disturbances. This paper summarizes the planning of the experiment, the instruction for the experiment and logs, the data recording conditions, recorded signal wave forms and the results of power spectral analysis. (author)

  15. Reactor dynamics experiment of nuclear ship Mutsu using pseudo random signal (III). The third experiment

    International Nuclear Information System (INIS)

    Hayashi, Koji; Shimazaki, Junya; Nabeshima, Kunihiko; Ochiai, Masaaki; Shinohara, Yoshikuni; Inoue, Kimihiko.

    1995-03-01

    In order to investigate dynamics of the reactor plant of the nuclear ship Mutsu, the third reactor noise experiment using pseudo random binary sequences (PRBS) was performed on September 16, 1991 in the third experimental navigation. The experiments using both reactivity and load disturbances were performed at 70% of reactor power and under a normal sea condition. Each PRBS was applied by manual operation of the control rod or the main steam valve. Various signals of the plant responses and of the acceleration of ship motion were measured. Furthermore, natural reactor noise signals were measured after each PRBS experiment in order to evaluate the effects of the PRBS disturbances. This paper summarizes the planning of the experiment, the instruction for the experiment and logs, the data recording conditions, recorded signal wave forms and the results of power spectral analysis. (author)

  16. Complementarity of integral and differential experiments for reactor physics purposes

    International Nuclear Information System (INIS)

    Tellier, Henry.

    1981-04-01

    In this paper, the following topics are studied: uranium 238 effective integral; thermal range uranium 238 capture cross section; Americium 242 m capture cross section. The mentioned examples show that differential and integral experiments are both useful to the reactor physicists

  17. French gas cooled reactor experience with moisture ingress

    International Nuclear Information System (INIS)

    Bastien, D.; Brie, M.

    1995-01-01

    During the history of operation of six gas cooled reactors in France, some experience has been gained with accidental water ingress into the primary system. This occurred as a result of leaks in steam generators. This paper describes the cause of the leaks, and the resulting consequences. (author). 2 refs, 8 figs

  18. Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    T. A. Tomberlin; S. B. Grover

    2004-11-01

    The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment.

  19. Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    T. A. Tomberlin; S. B. Grover

    2004-01-01

    The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment

  20. Review of experiments for research reactors - approved 1974

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    This standard establishes guidelines for the review and approval of experiments performed at research reactor facilities. This standard identifies the major areas that shall be reviewed for each experiment to ensure that it (a) falls within the limits delineated in the technical specifications, (b) does not present an unreviewed safety question as defined in 10 CFR Section 50.59 π2-, (c) does not constitute a threat to the health and safety of any individual or group of individuals, and (d) does not constitute a hazard to the reactor facility or other equipment. In addition, this standard recommends a system for classifying experiments to establish levels of review and approval commensurate with the level of risk inherent in the experiment

  1. Estimation of radioactivity in structural materials of ETRR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Imam, M [National Center for Nuclear Safety and Radiation Control Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    Precise knowledge of the thermal neutron flux in the different structural materials of a reactor is necessary to estimate the radioactive inventory in these materials that are needed in any decommissioning study of the reactor. ETRR-1 is a research reactor that went critical on 2/1691. In spite of this long age of the reactor, the effective operation time of this reactor is very short since the reactor was shutdown for long periods. Because of this long age one may think of reactor decommissioning. For this purpose, the radioactivity of the reactor structural materials was estimated. Apart from the reactor core, the important structural materials in the ETRR-1 are the reactor tank, shielding concrete, and the graphite thermal column. The thermal neutron flux was determined by the monte Carlo method in these materials and the isotope inventory and the radioactivity were calculated by the international code ORIGEN-JR. 1 fig.

  2. Packaging and shipment of U.S. breeder reactor experiments

    International Nuclear Information System (INIS)

    Berger, J.D.

    1980-01-01

    Irradiation testing of fuels and materials in the Fast Test Reactor (FTR) required development of a shipping cask (designated T-3) and associated hardware for loading and shipping of these experiments to postirradiation examination facilities. The T-3 shipping-cask program included design, fabrication, and testing of internal cask packages to protect the experiments during loading, shipping, and unloading. The cask was designed for loading in both the vertical and horizontal attitudes

  3. Combined potential of future long-baseline and reactor experiments

    International Nuclear Information System (INIS)

    Huber, P.; Lindner, M.; Rolinec, M.; Schwetz, T.; Winter, W.

    2005-01-01

    We investigate the determination of neutrino oscillation parameters by experiments within the next ten years. The potential of conventional beam experiments (MINOS, ICARUS, OPERA), superbeam experiments (T2K, NOνA), and reactor experiments (D-CHOOZ) to improve the precision on the 'atmospheric' parameters Δm 31 2 , θ 23 , as well as the sensitivity to θ 13 are discussed. Further, we comment on the possibility to determine the leptonic CP-phase and the neutrino mass hierarchy if θ 13 turns out to be large

  4. Fission product behavior in the Molten Salt Reactor Experiment

    International Nuclear Information System (INIS)

    Compere, E.L.; Kirslis, S.S.; Bohlmann, E.G.; Blankenship, F.F.; Grimes, W.R.

    1975-10-01

    Essentially all the fission product data for numerous and varied samples taken during operation of the Molten Salt Reactor Experiment or as part of the examination of specimens removed after particular phases of operation are reported, together with the appropriate inventory or other basis of comparison, and relevant reactor parameters and conditions. Fission product behavior fell into distinct chemical groups. Evidence for fission product behavior during operation over a period of 26 months with 235 U fuel (more than 9000 effective full-power hours) was consistent with behavior during operation using 233 U fuel over a period of about 15 months (more than 5100 effective full-power hours)

  5. Experiments in power distribution control on the IRT-2000 reactor

    International Nuclear Information System (INIS)

    Filipchuk, E.V.; Potapenko, P.T.; Trofimov, A.P.; Kosilov, A.N.; Neboyan, V.T.; Timokhin, E.S.

    1975-01-01

    The results from the experimental investigations of a system for regulating the neutron field on a research reactor IRT-2000 are shown. The right of such experiments on a reactor with a little active zone is substantiated. A successful attempt was made in this work to apply primary elements of straight charging in the neutron field regulating system. A system with independent instrumentally local regulators, a system with hard cross connections and a structure with a ''floating'' installation are studied. Serial common industrial regulators BRT-2 were used

  6. Decommissioning experience of the Japan power demonstration reactor

    International Nuclear Information System (INIS)

    Hoshi, T.; Yanagihara, S.; Tachibana, M.; Momma, T.

    1992-01-01

    Actual dismantling of the Japan Power Demonstration Reactor (JPDR) has been progressing since 1986 aiming to make stage 3 condition as the final goal. Such highly activated components as the reactor pressure vessel (RPV) and the inner portion of biological shield concrete close to the RPV have removed using the remotely operated cutting machines. Useful data on the dismantling techniques and their safety as well as the manpower expenditure and radiation exposure of workers have been obtained. Experiences gained through the dismantling works are described in this paper. (author)

  7. Training experience at Experimental Breeder Reactor II

    International Nuclear Information System (INIS)

    Driscoll, J.W.; McCormick, R.P.; McCreery, H.I.

    1978-01-01

    The EBR-II Training Group develops, maintains,and oversees training programs and activities associated with the EBR-II Project. The group originally spent all its time on EBR-II plant-operations training, but has gradually spread its work into other areas. These other areas of training now include mechanical maintenance, fuel manufacturing facility, instrumentation and control, fissile fuel handling, and emergency activities. This report describes each of the programs and gives a statistical breakdown of the time spent by the Training Group for each program. The major training programs for the EBR-II Project are presented by multimedia methods at a pace controlled by the student. The Training Group has much experience in the use of audio-visual techniques and equipment, including video-tapes, 35 mm slides, Super 8 and 16 mm film, models, and filmstrips. The effectiveness of these techniques is evaluated in this report

  8. International Reactor Physics Experiment Evaluation (IRPhE) Project

    International Nuclear Information System (INIS)

    2013-01-01

    The International Reactor Physics Experiment Evaluation (IRPhE) Project aims to provide the nuclear community with qualified benchmark data sets by collecting reactor physics experimental data from nuclear facilities, worldwide. More specifically the objectives of the expert group are as follows: - maintaining an inventory of the experiments that have been carried out and documented; - archiving the primary documents and data released in computer-readable form; - promoting the use of the format and methods developed and seek to have them adopted as a standard. For those experiments where interest and priority is expressed by member countries or working parties and executive groups within the NEA provide guidance or co-ordination in: - compiling experiments into a standard international agreed format; - verifying the data, to the extent possible, by reviewing original and subsequently revised documentation, and by consulting with the experimenters or individuals who are familiar with the experimenters or the experimental facility; - analysing and interpreting the experiments with current state-of-the-art methods; - publishing electronically the benchmark evaluations. The expert group will: - identify gaps in data and provide guidance on priorities for future experiments; - involve the young generation (Masters and PhD students and young researchers) to find an effective way of transferring know-how in experimental techniques and analysis methods; - provide a tool for improved exploitation of completed experiments for Generation IV reactors; - coordinate closely its work with other NSC experimental work groups in particular the International Criticality Safety Benchmark Evaluation Project (ICSBEP), the Shielding Integral Benchmark Experiment Data Base (SINBAD) and others, e.g. knowledge preservation in fast reactors of the IAEA, the ANS Joint Benchmark Activities; - keep a close link with the working parties on scientific issues of reactor systems (WPRS), the expert

  9. Latest Results from the Daya Bay Reactor Neutrino Experiment

    CERN Multimedia

    CERN. Geneva

    2014-01-01

    Among all the fundamental particles that have been experimentally observed, neutrinos remain one of the least understood. The Daya Bay Reactor Neutrino Experiment in China consists of eight identical detectors placed underground at different baselines from three groups of nuclear reactors, a configuration that is ideally suited for studying the properties of these elusive particles. This talk will present three sets of results that have just recently been released by the Daya Bay Collaboration: (i) a precision measurement of the oscillation parameters that drive the disappearance of electron antineutrinos at short baselines, (ii) a search for sterile neutrino mixing, and (iii) a high-statistics determination of the absolute flux and spectrum of reactor-produced electron antineutrinos. All of these results extend the limits of our knowledge in their respective areas and thus shed new light on neutrinos and the physics that surround them.

  10. Determination flux in the Reactor JEN-1

    International Nuclear Information System (INIS)

    Manas Diaz, L.; Montes Ponce de leon, J.

    1960-01-01

    This report summarized several irradiations that have been made to determine the neutron flux distributions in the core of the JEN-1 reactor. Gold foils of 380 μ gr and Mn-Ni (12% de Ni) of 30 mg have been employed. the epithermal flux has been determined by mean of the Cd radio. The resonance integral values given by Macklin and Pomerance have been used. (Author) 9 refs

  11. Experiment operations plan for the MT-4 experiment in the NRU reactor

    International Nuclear Information System (INIS)

    Russcher, G.E.; Wilson, C.L.; Parchen, L.J.; Marshall, R.K.; Hesson, G.M.; Webb, B.J.; Freshley, M.D.

    1983-06-01

    A series of thermal-hydraulic and cladding materials deformation experiments were conducted using light-water reactor fuel bundles as part of the Pacific Northwest Laboratory Loss-of-Coolant Accident (LOCA) Simulation Program. This report is the formal operations plan for MT-4 - the fourth materials deformation experiment conducted in the National Research Universal (NRU) reactor, Chalk River, Ontario, Canada. A major objective of MT-4 was to simulate a pressurized water reactor LOCA that could induce fuel rod cladding deformation and rupture due to a short-term adiabatic transient and a peak fuel cladding temperature of 1200K (1700 0 F)

  12. Chasing {theta}{sub 13} with new reactor neutrino experiments

    Energy Technology Data Exchange (ETDEWEB)

    Lasserre, Th. [DSM/DAPNIA/SPP, CEA/Saclay, 91191 Gif-sur-Yvette (France)

    2005-12-15

    It is now widely accepted that a new middle baseline disappearance reactor neutrino experiment with multiple detectors could provide a clean measurement of the {theta}{sub 13} mixing angle, free from any parameter degeneracies and correlations induced by matter effect and the unknown leptonic Dirac CP phase. The current best constraint on the third mixing angle comes from the Chooz reactor neutrino experiment sin{sup 2}(2{theta}{sub 13})<0.2 (90 % C.L., {delta}m{sub atm}{sup 2}=2.010{sup -3} eV{sup 2}). Several projects of experiment, with different timescales, have been proposed over the last two years all around the world. Their sensitivities range from sin{sup 2}(2{theta}{sub 13})<0.01 to 0.03, having thus an excellent discovery potential of the {nu}{sub e} fraction of {nu}{sub 3}.

  13. Operating Experience with Power Reactors. Proceedings of the Conference on Operating Experience with Power Reactors. Vol. I

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1963-10-15

    At the beginning of 1963 nuclear power plants produced some 3 500 000kW of electrical power to different distribution grids around the world. Much significant operating experience has been gained with these power reactors, but this experience is often not collected in such a way as to make it easily available. The International Atomic Energy Agency convened a Conference on Operating Experience with Power Reactors in Vienna from 4-8 June 1963 which was attended by 240 participants representing 27 of the Agency's Member States and six international organizations. At the Conference, 42 papers giving detailed experience with more than 20 nuclear power stations were discussed. Although similar meetings on a national or regional scale have been held earlier in various countries, this is the first arranged by the Agency on a world-wide basis. Some of the detailed material may have been given earlier but for the most part it represents new and recently acquired experience, and for the first time it has been possible to compile in one place such extensive material on the operating experience with power reactors. The Conference discussed the experience gained both generally in the context of national and international nuclear power development programmes, and more specifically in the detailed operating experience with different power reactor stations. In addition, various plant components, fuel cycles, staffing of nuclear plants and licensing of such staff were treated. It is hoped that these Proceedings will be of interest not only to nuclear plant designers and operators who daily encounter problems similar to those discussed by the Conference, but also to those guiding the planning and implementation of power development programmes.

  14. Operating Experience with Power Reactors. Proceedings of the Conference on Operating Experience with Power Reactors. Vol. II

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1963-10-15

    At the beginning of 1963 nuclear power plants produced some 3 500 000 kW of electrical power to different distribution grids around the world. Much significant operating experience has been gained with these power reactors, but this experience is often not collected in such a way as to make it easily available. The International Atomic Energy Agency convened a Conference on Operating Experience with Power Reactors in Vienna from 4 -8 June 1963 which was attended by 240 participants representing 27 of the Agency's Member States and six international organizations. At the Conference, 42 papers giving detailed experience with more than 20 nuclear power stations were discussed. Although similar meetings on a national or regional scale have been held earlier in various countries, this is the first arranged by the Agency on a world-wide basis. Some of the detailed material may have been given earlier but for the most part it represents new and recently acquired experience, and for the first time it has been possible to compile in one place such extensive material on the operating experience with power reactors. The Conference discussed the experience gained both generally in the context of national and international nuclear power development programmes, and more specifically in the detailed operating experience with different power reactor stations. In addition, various plant components, fuel cycles, staffing of nuclear plants and licensing of such staff were treated. It is hoped that these Proceedings will be of interest not only to nuclear , plant designers and operators who daily encounter problems similar to those discussed by the Conference, but also to those guiding the planning and implementation of power development programmes.

  15. Field experience in use of radiation instruments in Cirus reactor

    International Nuclear Information System (INIS)

    Ramesh, N.; Sharma, R.C.; Agarwal, S.K.; Sawant, D.K.; Yadav, R.K.B.; Prasad, S.K.

    2005-01-01

    Cirus, located at Bhabha Atomic Research Centre, is a 40 MW (Th) research reactor fuelled by natural uranium, moderated by heavy water and cooled by de-mineralized light water. Graphite is used as reflector in this reactor. The reactor, commissioned in the year 1960, was in operation with availability factor of about 70% till early nineties. There after signs of ageing started surfacing up. After ageing studies, refurbishment plan was finalized and executed during the period from 1997-2002. after successful refurbishment, the reactor is in operation at full power. A wide range of radiation instruments have been used at Cirus for online monitoring of the radiological status of various process systems and environmental releases. Also, variety of survey meters, counting systems and monitors have been used by the health physics unit of the reactor for radiation hazard control. Many of these instruments, which were originally of Canadian design, have undergone changes due to obsolescence or as part of upgradation. This paper describes the experience with the radiation instruments of Cirus, bringing out their effectiveness in meeting the design intent, difficulties faced in their field use, and modifications carried out based on the performance feed back. Also, this paper highlights the areas where further efforts are needed to develop nuclear instrumentation to further strengthen monitoring and surveillance. (author)

  16. Nuclear data and integral experiments in reactor physics

    International Nuclear Information System (INIS)

    Farinelli, U.

    1980-01-01

    The material given here broadly covers the content of the 10 lectures delivered at the Winter Course on Reactor Theory and Power Reactors, ICTP, Trieste (13 February - 10 March 1978). However, the parts that could easily be found in the current literature have been omitted and replaced with the appropriate references. The needs for reactor physics calculations, particularly as applicable to commercial reactors, are reviewed in the introduction. The relative merits and shortcomings of fundamental and semi-empirical methods are discussed. The relative importance of different nuclear data, the ways in which they can be measured or calculated, and the sources of information on measured and evaluated data are briefly reviewed. The various approaches to the condensation of nuclear data to multigroup cross sections are described. After some consideration to the sensitivity calculations and the evaluation of errors, some of the most important type of integral experiments in reactor physics are introduced, with a view to showing the main difficulties in the interpretation and utilization of their results and the most recent trends in experimentation. The conclusions try to assign some priorities in the implementation of experimental and calculational capabilities, especially for a developing country. (author)

  17. Integral tightness measurements at the Paks-1 nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Taubner, R.; Techy, Z. (Villamosenergiaipari Kutato Intezet, Budapest (Hungary))

    1983-01-01

    The containment system experiments of the Paks-1 nuclear reactor are described. The integrated tightness measurements of the hermetic system were completed in 1982. The principles and methods and the evaluation of the results of the measurements are discussed. Some features of the filtration characteristics are demonstrated using relative values and a method enabling the description of the physical contents of the characteristics by flow technical functions is outlined.

  18. Experiments prior to construction of the Rapsodie reactor (1962)

    International Nuclear Information System (INIS)

    Vautrey, L.; Zaleski, C.P.

    1962-01-01

    Before proceeding to the construction of the various reactor components described in the paper 'Fast Breeder Reactor Rapsodie', many experimental studies of a hydraulic, thermal and mechanical character have been carried out, or are under consideration, to test the validity of the principles adopted in the Preliminary Project. This paper deals with the most important of these: 1. Studies of coolant circuit components: sodium pumps (mechanical or electromagnetic), Na-NaK and NaK ir heat exchangers, measuring instruments (flow rates, temperatures), sodium purification circuits, etc. 2. Studies in cooling of fuel and fertile assemblies: a) study of the sodium cooling carried out by means of hydraulic mockups (scale of 1: 1 or over) reproducing the flow of the coolant fluid in the piping, upstream from and inside the fuel and fertile elements. b) study of the cooling by gas and by immersion in lead, employed during handling and storage operations. 3. Studies of special reactor devices: fusible rotating linkage, parts of the control rod mechanisms. 4. Study of the reactor block and coolant circuits as a whole. This study is to begin at the end of the year. The mock-up, now nearing completion, reproduces on a scale of 1: 1 the installation provided in the Preliminary Project and includes: the reactor block, to which is connected a high flow ate sodium circuit, permitting of long-term tests and thermal shocks, and also, a control rod testing circuit; complete installation of the 1 MW and 10 MW coolant circuits, the performances of which it will be possible to check under various operational conditions. 5. A safety study carried out on a 3: 10 scale mock p comprising the whole of the reactor block and shielding, with the object of limiting the effects of any accidental liberation of energy of an explosive character. (authors) [fr

  19. Stability analysis of the Ghana Research Reactor-1 (GHARR-1)

    International Nuclear Information System (INIS)

    Della, R.; Alhassan, E.; Adoo, N.A.; Bansah, C.Y.; Nyarko, B.J.B.; Akaho, E.H.K.

    2013-01-01

    Highlights: • We developed a theoretical model to study the stability of the Ghana Research Reactor-1. • The neutronics transfer function was described by the point kinetics model for a single group of delayed neutrons. • The thermal hydraulics transfer function was based on the modified lumped parameter concept. • A computer code, RESA (REactor Stability Analysis) was developed. • Results show that the closed-loop transfer function was stable and well damped for variable operating power levels. - Abstract: A theoretical model has been developed to study the stability of the Ghana Research Reactor one (GHARR-1). The closed-loop transfer function of GHARR-1 was established based on the model, which involved the neutronics and the thermal hydraulics transfer functions. The reactor kinetics was described by the point kinetics model for a single group of delayed neutrons, whilst the thermal hydraulics transfer function was based on the modified lumped parameter concept. The inherent internal feedback effect due to the fuel and the coolant was represented by the fuel temperature coefficient and the moderator temperature coefficient respectively. A computer code, RESA (REactor Stability Analysis), entirely in Java was developed based on the model for systems analysis. Stability analysis of the open-loop transfer function of GHARR-1 based on the Nyquist criterion and Bode diagrams using RESA, has shown that the closed-loop transfer function was marginally stable for variable operating power levels. The relative stability margins of GHARR-1 were also identified

  20. Quality assurance plan for the Molten Salt Reactor Experiment Remediation Project at the Oak Ridge National Laboratory. Phase 1 -- Interim corrective measures and Phase 2 -- Purge and trap reactive gases

    International Nuclear Information System (INIS)

    1995-11-01

    This Quality Assurance Plan (QAP) identifies and describes the systems utilized by the Molten Salt Reactor Experiment Remediation Project (MSRERP) personnel to implement the requirements and associated applicable guidance contained in the Quality Program Description Y/QD-15 Rev. 2 (Energy Systems 1995f). This QAP defines the quality assurance (QA) requirements applicable to all activities and operations in and directly pertinent to the MSRERP Phase 1--Interim Corrective Measures and Phase 2--Purge and Trap objectives. This QAP will be reviewed, revised, and approved as necessary for Phase 3 and Phase 4 activities. This QAP identifies and describes the QA activities and procedures implemented by the various Oak Ridge National Laboratory support organizations and personnel to provide confidence that these activities meet the requirements of this project. Specific support organization (Division) quality requirements, including the degree of implementation of each, are contained in the appendixes of this plan

  1. UK experience of safety requirements for thermal reactor stations

    International Nuclear Information System (INIS)

    Matthews, R.R.; Dale, G.C.; Tweedy, J.N.

    1977-01-01

    The paper summarises the development of safety requirements since the first of the Generating Boards' Magnox reactors commenced operation in 1962 and includes A.G.R. safety together with the preparation of S.G.H.W.R. design safety criteria. It outlines the basic principles originally adopted and shows how safety assessment is a continuing process throughout the life of a reactor. Some description is given of the continuous effort over the years to obtain increased safety margins for existing and new reactors, taking into account the construction and operating experience, experimental information, and more sophisticated computer-aided design techniques which have become available. The main safeguards against risks arising from the Generating Boards' reactors are the achievement of high standards of design, construction and operation, in conjunction with comprehensive fault analyses to ensure that adequate protective equipment is provided. The most important analyses refer to faults which can lead to excessive fuel element temperatures arising from an increase in power or a reduction in cooling capacity. They include the possibility of unintended control rod withdrawal at power or at start-up, coolant flow failure, pressure circuit failure, loss of boiler feed water, and failure of electric power. The paper reviews the protective equipment, and the policy for reactor safety assessments which include application of maximum credible accident philosophy and later the limited use of reliability and probability methods. Some of the Generating Boards' reactors are now more than half way through their planned working lives and during this time safety protective equipment has occasionally been brought into operation, often for spurious reasons. The general performance, of safety equipment is reviewed particularly for incidents such as main turbo-alternator trip, circulator failure, fuel element failures and other similar events, and some problems which have given rise to

  2. Fast breeder reactors: Experience and trends. V. 2

    International Nuclear Information System (INIS)

    1986-01-01

    The IAEA Symposium on ''Fast Breeder Reactors: Experience and Future Trends'' was held, at the invitation of the Government of France, in Lyons, France, on 22-26 July 1985. It was hosted by the French Commissariat a l'energie atomique and Electricite de France. The purpose of the Symposium was to review the experience gained so far in the field of LMFBRs, taking into account the constructional, operational, technological, economic and fuel cycle aspects, and to consider the developmental trends as well as the international co-operation in fast breeder reactor design and utilization. The Symposium was attended by almost 400 participants (340 participants, 35 observers and 20 journalists) from 25 countries and five international organizations. More than 80 papers were presented and discussed during six regular sessions and four poster sessions. A separate abstract was prepared for each of these papers

  3. Reactor inventory monitoring system for Angra-1 reactor

    International Nuclear Information System (INIS)

    S Neto, Joaquim A.; Silva, Marcos C.; Pinheiro, Ronaldo F.M.; Soares, Milton; Martinez, Aquilino; Comerlato, Cesar A.; Oliveira, Eugenio A.

    1996-01-01

    This work describes the project of Reactor Inventory Monitoring System, which will be installed in Angra I Nuclear Power Plant. The inventory information is important to the operators take corrective actions in case of an incident that may cause a failure in the core cooling. (author)

  4. Slow control systems of the Reactor Experiment for Neutrino Oscillation

    International Nuclear Information System (INIS)

    Choi, J.H.; Jang, H.I.; Choi, W.Q.; Choi, Y.; Jang, J.S.; Jeon, E.J.; Joo, K.K.; Kim, B.R.; Kim, H.S.; Kim, J.Y.; Kim, S.B.; Kim, S.Y.; Kim, W.; Kim, Y.D.; Ko, Y.J.; Lee, J.K.; Lim, I.T.; Pac, M.Y.; Park, I.G.; Park, J.S.

    2016-01-01

    The RENO experiment has been in operation since August 2011 to measure reactor antineutrino disappearance using identical near and far detectors. For accurate measurements of neutrino mixing parameters and efficient data taking, it is crucial to monitor and control the detector in real time. Environmental conditions also need to be monitored for stable operation of detectors as well as for safety reasons. In this paper, we report the design, hardware, operation, and performance of the slow control system.

  5. Trends and experiences in reactor coolant pump motors

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    A review of the requirements and features of these motors is given as background along with a discussion of trends and experiences. Included are a discussion of thrust bearings and a review of safety related requirements and design features. Primary coolant pump motors are vertical induction motors for pumps that circulate huge quantities of water through the reactor core to carry the heat generated there to steam generator heat exchangers. 4 refs

  6. Experiments and analysis of thermal stresses around the nozzle of the reactor vessel

    International Nuclear Information System (INIS)

    Song, D.H.; Oh, J.H.; Song, H.K.; Park, D.S.; Shon, K.H.

    1981-01-01

    This report describes the results of analysis and experiments on the thermal stress around the reactor vessel nozzle performed to establish a capability of thermal stress analysis of pressure vessel subjected to thermal loadings. Firstly, heat conduction analysis during reactor design transients and analysis on the experimental model were performed using computer code FETEM-1 for the purpose of verification of FETEM-1 which was developed in 1979 and will be used to obtain the temperature distribution in a solid body under the steady-state and the transient conditions. The results of the analysis was compared to the results in the Stress Report of Kori-1 reactor vessel and those from experiments on the model, respectively

  7. Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Bess, J. D.; Briggs, J. B.; Gulliford, J.; Ivanova, T.; Rozhikhin, E. V.; Semenov, M. Yu.; Tsibulya, A. M.; Koscheev, V. N.

    2017-07-01

    Overview of Experiments to Study the Physics of Fast Reactors Represented in the International Directories of Critical and Reactor Experiments John D. Bess Idaho National Laboratory Jim Gulliford, Tatiana Ivanova Nuclear Energy Agency of the Organisation for Economic Cooperation and Development E.V.Rozhikhin, M.Yu.Sem?nov, A.M.Tsibulya Institute of Physics and Power Engineering The study the physics of fast reactors traditionally used the experiments presented in the manual labor of the Working Group on Evaluation of sections CSEWG (ENDF-202) issued by the Brookhaven National Laboratory in 1974. This handbook presents simplified homogeneous model experiments with relevant experimental data, as amended. The Nuclear Energy Agency of the Organization for Economic Cooperation and Development coordinates the activities of two international projects on the collection, evaluation and documentation of experimental data - the International Project on the assessment of critical experiments (1994) and the International Project on the assessment of reactor experiments (since 2005). The result of the activities of these projects are replenished every year, an international directory of critical (ICSBEP Handbook) and reactor (IRPhEP Handbook) experiments. The handbooks present detailed models of experiments with minimal amendments. Such models are of particular interest in terms of the settlements modern programs. The directories contain a large number of experiments which are suitable for the study of physics of fast reactors. Many of these experiments were performed at specialized critical stands, such as BFS (Russia), ZPR and ZPPR (USA), the ZEBRA (UK) and the experimental reactor JOYO (Japan), FFTF (USA). Other experiments, such as compact metal assembly, is also of interest in terms of the physics of fast reactors, they have been carried out on the universal critical stands in Russian institutes (VNIITF and VNIIEF) and the US (LANL, LLNL, and others.). Also worth mentioning

  8. Heavy reflector experiments in the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    Santos, Adimir dos; Silva, Graciete Simoes de Andrade e; Mura, Luis Felipe; Fuga, Rinaldo; Jerez, Rogerio; Mendonca, Arlindo Gilson

    2012-01-01

    Full text: The heavy reflector experiments performed in the IPEN/MB-01 research reactor facility comprise a set of critical configurations employing the standard 28x26-fuel-rod configuration. The heavy reflector either Stainless Steel, Carbon Steel or Nickel plates was placed at one of the faces of the IPEN/MB-01 reactor. Criticality is achieved by inserting the control banks BC1 and BC2 to the critical position. 32 plates around 0.3 mm thick were used in the experiment. The chosen distance between last fuel rod row and the first laminate for both type of laminates was 5.5 mm. Considering initially the SS case, the experimental data reveal that the reactivity decreases up to the sixth plate and after that it increases, becomes nearly zero (which was equivalent to initial zero excess reactivity with zero plates) for the 21 plates case and reaches a value of 154.91 pcm when the whole set of 32 plates are inserted in the reflector. This is a very striking result because it demonstrates that when all 32 plates are inserted in the reflector there is a net gain of reactivity. The reactivity behavior demonstrates all the physics events already mentioned in this work. When the number of plates are small (around 6), the neutron absorption in the plates is more important than the neutron reflection and the reactivity decreases. This condition holds up to a point where the neutron reflection becomes more important than the neutron absorption in the plates and the reactivity increases. The experimental data for the Carbon Steel and Nickel case shows the main features of the SS case, but for the Carbon Steel case the reactivity gain is small, thus demonstrating that Carbon Steel or essentially iron has not the reflector capability as the SS laminates do. The measured data of Nickel plates show a higher reactivity gain, thus demonstrating that Nickel is a good reflector. The theoretical analysis employing MCNP5 and ENDF/B-VII.0 show that the SS calculated results are in a good

  9. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Grover, S. Blaine

    2009-01-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  10. Thermal hydraulic and safety analyses for Pakistan Research Reactor-1

    International Nuclear Information System (INIS)

    Bokhari, I.H.; Israr, M.; Pervez, S.

    1999-01-01

    Thermal hydraulic and safety analysis of Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel have been performed using computer code PARET. The present core comprises of 29 standard and 5 control fuel elements. Results of the thermal hydraulic analysis show that the core can be operated at a steady-state power level of 10 MW for a flow rate of 950 m 3 /h, with sufficient safety margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions. The events studied include: start-up accident; accidental drop of a fuel element in the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is therefore concluded that the reactor can be safely operated at 10 MW without compromising safety. (author)

  11. Operating experiences since rise-to-power test in high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Watanabe, Shuji; Motegi, Toshihiro; Kawano, Shuichi; Kameyama, Yasuhiko; Sekita, Kenji; Kawasaki, Kozo

    2007-03-01

    The rise-to-power test of the High Temperature Engineering Test Reactor (HTTR) was actually started in April 2000. The rated thermal power of 30MW and the rated reactor outlet coolant temperature of 850degC were achieved in the middle of Dec. 2001. After that, the reactor thermal power of 30MW and the reactor outlet coolant temperature of 950degC were achieved in the final rise-to-power test in April 2004. After receiving the operation licensing at 850degC, the safety demonstration tests have conducted to demonstrate inherent safety features of the HTGRs as well as to obtain the core and plant transient data for validation of safety analysis codes and for establishment of safety design and evaluation technologies. This paper summarizes the HTTR operating experiences for six years from start of the rise-to-power test that are categorized into (1) Operating experiences related to advanced gas-cooled reactor design, (2) Operating experiences for improvement of the performance, (3) Operating experiences due to fail of system and components. (author)

  12. Self-sustaining nuclear pumped laser-fusion reactor experiment

    International Nuclear Information System (INIS)

    Boody, F.P.; Choi, C.K.; Miley, G.H.

    1977-01-01

    The features of a neutron feedback nuclear pumped (NFNP) laser-fusion reactor equipment were studied with the intention of establishing the feasibility of the concept. The NFNP laser-fusion concept is compared schematically to electrically pumped laser fusion. The study showed that, once a method of energy storage has been demonstrated, a self-sustaining fusion-fission hybrid reactor with a ''blanket multiplication'' of two would be feasible using nuclear pumped Xe F* excimer lasers having efficiencies of 1 to 2 percent and D-D-T pellets with gains of 50 to 100

  13. Light water reactor mixed-oxide fuel irradiation experiment

    International Nuclear Information System (INIS)

    Hodge, S.A.; Cowell, B.S.; Chang, G.S.; Ryskamp, J.M.

    1998-01-01

    The United States Department of Energy Office of Fissile Materials Disposition is sponsoring and Oak Ridge National Laboratory (ORNL) is leading an irradiation experiment to test mixed uranium-plutonium oxide (MOX) fuel made from weapons-grade (WG) plutonium. In this multiyear program, sealed capsules containing MOX fuel pellets fabricated at Los Alamos National Laboratory (LANL) are being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The planned experiments will investigate the utilization of dry-processed plutonium, the effects of WG plutonium isotopics on MOX performance, and any material interactions of gallium with Zircaloy cladding

  14. Measurements of reactivity of reactor G1

    International Nuclear Information System (INIS)

    Bernot, J.; Koechlin, J.C.; Portes, L.; Teste du Bailler, A.

    1957-01-01

    The various methods used during the physical study of the reactor G1 to determine the variations of the effective multiplication factor consecutive to a given change in the geometry of the multiplying medium, are presented and discussed. The comparison of the results obtained by these various methods has allowed their validity to be tested and precise conditions of use to be given. In the first part are presented the principles used and their ranges of validity. In the second part the experimental results are given, together with some indications on their comparison with theoretical estimations. (author) [fr

  15. Education and research at the VR-1 Vrabec training reactor facility

    International Nuclear Information System (INIS)

    Matejka, K.

    1993-01-01

    The results of 12 years' efforts devoted to the construction of the VR-1 ''Vrabec'' training reactor at the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague and to establishing the training reactor department, as well as the contribution of the training reactor facility to the teaching and scientific activities of the Faculty are presented in a comprehensive manner. The thesis is divided into 2 parts: (i) preconditions, reactor construction and commissioning, and constituting the reactor department, and (ii) basic and comprehensive information concerning the current utilization of the reactor for the benefit of students from various university level institutions. The prospects of scientific activities of the department are also outlined. Attention is paid to selected nuclear safety aspects of the reactor during operation and teaching of students, as well as to its innovated digital control system whose implementation is planned. The results achieved are compared with the initial goals and with similar experience abroad. (P.A.)

  16. Complementary role of critical integral experiment and power reactor start-up experiments for LMFBR neutronics data and method validation

    International Nuclear Information System (INIS)

    Salvatores, M.

    1986-09-01

    Both critical experiments and power reactor results play at present a complementary role in reducing the uncertainties in Key design parameters for LMFBR, which can be relevant for the economic performances of this type of reactors

  17. Status of the Daya Bay Reactor Neutrino Oscillation Experiment

    International Nuclear Information System (INIS)

    Lin, Cheng-Ju Stephen

    2010-01-01

    The last unknown neutrino mixing angle θ 13 is one of the fundamental parameters of nature; it is also a crucial parameter for determining the sensitivity of future long-baseline experiments aimed to study CP violation in the neutrino sector. Daya Bay is a reactor neutrino oscillation experiment designed to achieve a sensitivity on the value of sin 2 (2*θ 13 ) to better than 0.01 at 90% CL. The experiment consists of multiple identical detectors placed underground at different baselines to minimize systematic errors and suppress cosmogenic backgrounds. With the baseline design, the expected anti-neutrino signal at the far site is about 360 events per day and at each of the near sites is about 1500 events per day. An overview and current status of the experiment will be presented.

  18. Reactor Neutrino Detection for Non Proliferation with the NUCIFER Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Bouvet, L. [CEA, Centre de Saclay, IRFU, Gif-sur-Yvette, (France); Bouvier, S.; Bui, V. M. [Laboratoire Subatech, Ecole des Mines, Nantes Cedex 3 (France); others, and

    2012-06-15

    Neutrinos are the most abundant matter particles in the Universe. Thoroughly investigated in basic science, the neutrino field is now delivering first applications to the monitoring of nuclear reactors. The neutrinos are emitted in the decay chain of the fission products; therefore measuring their flux provides real-time information, directly related to the fission process occurring in the reactor core. Because of the very weak interaction of neutrinos with matter a neutrino detector can stand outside the core containment vessel and provide a non-intrusive and inherently tamper resistant measurement. After a brief review of the existing data and worldwide projects, we present the NUCIFER experiment. The active part of the detector is a tank filled up with one ton of Gadolinium-doped liquid scintillator. Sixteen photomultiplier tubes, isolated from the liquid by an acrylic buffer, read out the light produced by the interaction of a neutrino with the protons of the liquid. The tank is surrounded by plastic scintillator plates to veto the cosmic rays. Then polyethylene and lead shielding suppress the background coming from external neutrons and gamma rays respectively. The NUCIFER detector has been designed for an optimal compromise between the detection performances and the specifications of operation in a safeguards regime. Its global footprint is 2.8 m x 2.8 m and it can monitor remotely the nuclear power plant thermal power and Plutonium content with very little maintenance on years scale. The experiment is currently installed near the OSIRIS research reactor (70 MWth) at CEA, in Saclay, France. First data are expected by May 2012. This work is done in contact with the IAEA/SGTN division that is currently investigating the potentiality of neutrinos as a novel safeguards tool. A dedicated working group has been created in 2010 to coordinate the simulation effort of various reactor types as well as the development of dedicated detectors and define and eventually

  19. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    International Nuclear Information System (INIS)

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor

  20. Final report on in-reactor creep-fatigue deformation behaviour of a CuCrZr alloy: COFAT 1

    DEFF Research Database (Denmark)

    Singh, Bachu Narain; Tähtinen, S.; Moilanen, P.

    CrZr(HT1) alloy exposed concurrently to flux of neutrons and creep-fatigue cyclic loading directly in a fission reactor. Special experimental facilities were designed and fabricated for this purpose. A number of in-reactor creep-fatigue experiments were successfully carried out in the BR-2 reactor at Mol...

  1. Leak-before-break experience in CANDU reactors

    International Nuclear Information System (INIS)

    Price, E.G.; Moan, G.D.; Coleman, C.E.

    1988-01-01

    In the Canada deuterium uranium (CANDU) reactor, each of the ∼ 400 hot pressure tubes containing the fuel bundles and the pressurized heat transport water is surrounded and insulated from the cold moderator by a calandria tube. The pressure tubes are made from cold-worked Zr-2.5 Nb with a minimum wall thickness of 4.19 mm, and the calandria tubes are made from annealed Zircaloy-2 with a minimum wall thickness of 1.37 mm. The annulus between these two tubes contains an inert gas. Leak-before-break has developed into an operational tool in CANDU reactors to prevent unstable failure of pressure tubes. A procedure for leak detection and reactor response has been developed from the use of the annulus gas, whose dew point is measured to ascertain if leaks have crept into the annulus. The characteristics of the crack are used to establish the response time for leak detection. The reactor is required to be shut down before the length of the slowly growing crack has reached the critical stage. This critical crack length, determined using slit burst tests on tubes, is the crack length at which the crack growth becomes unstable. The most likely crack growth mechanism is delayed hydride cracking. This mechanism requires three conditions to occur simultaneously: the material must be sensitive to delayed hydride cracking; zirconium hydrides must be present in the material; and the tensile stress must be sufficiently great

  2. Equipment for neutron measurements at VR-1 Sparrow training reactor

    International Nuclear Information System (INIS)

    Kolros, Antonin; Huml, Ondrej; Kos, Josef

    2008-01-01

    Full text: The VR-1 Sparrow training reactor is the experimental nuclear facility especially employed for education and teaching of students from different technical universities in the Czech Republic and other countries. Since 2005 the uniform all-purpose devices EMK310 have been used for measurement at reactor laboratory with different type of gas filled neutron detectors. The neutron detection system are employed for reactivity measurement, control rod calibration, critical experiment, study of delayed neutrons, study of nuclear reactor dynamics and study of detection systems dead time. The small dimension isotropic detectors are especially used for measurement of thermal neutron flux distribution inside the reactor core. The EMK-310 is a high performance, portable, three-channel fast amplitude analyzer designed for counting applications. It was developed for nuclear applications and made in close co-operation with firm TEMA Ltd. The precise rack eliminates electromagnetic disturbance and contains the control unit and four modules. The modules of high voltage supply and amplifier for gas filled detectors or scintillation probes are used in basic configuration. Software is tailored specifically to the reactor measurement and allows full online control. For applications involving the study of signals that may vary with the time, example study of delayed neutrons or nuclear reactor dynamics, the EMK-310 provides a Multichannel Scaling (MCS) acquisition mode. MCS dwell time can be set from 2 ms. Now, the new generation of digital multichannel analyzers DA310 is introduced. They have similarly attributes as EMK310 but the output information of unipolar signals from detector is more complete. The pipeline A/D converter with field programmable gate array (FPGA) is the hearth of the DA310 device. The resolution is 12 bits (4096 channels); the sample frequency is 80 MHz. The application for the neutron noise analysis is supposed. The correction method for non linearity

  3. Operation experience of the Indonesian multipurpose research reactor RSG-GAS

    Energy Technology Data Exchange (ETDEWEB)

    Hastowo, Hudi; Tarigan, Alim [Multipurpose Reactor Center, National Nuclear Energy Agency of the Republic of Indonesia (PRSG-BATAN), Kawasan PUSPIPTEK Serpong, Tangerang (Indonesia)

    1999-08-01

    RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U{sub 3}O{sub 8}-Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

  4. Operation experience of the Indonesian multipurpose research reactor RSG-GAS

    International Nuclear Information System (INIS)

    Hastowo, Hudi; Tarigan, Alim

    1999-01-01

    RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U 3 O 8 -Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

  5. Twelfth biennial conference on reactor operating experience maintenance and modifications for availability and efficiency

    International Nuclear Information System (INIS)

    1985-01-01

    Separate abstracts were prepared for the papers presented in the following topical sessions: (1) Formal Availability Improvement Programs; (2) Maintenance Aids and Diagnostics; (3) Nuclear Plant Inspection and Testing; (4) Secondary Cycle Improvement to Programs; (5) Recent Plant Experiences; (6) Training, Research, and Test Reactors; (7) Impact of Plant Maintenance on Availability; (8) Intragranular Stress Corrosion Cracking: Repair, Inspection, and Mitigating Techniques; (9) Availability Improvement - Recent Development and Trends; (10) Water Chemistry and Corrosion Control; and (11) ALARA: Experiences with Dose Reduction

  6. Validation study of the reactor physics lattice transport code WIMSD-5B by TRX and BAPL critical experiments of light water reactors

    International Nuclear Information System (INIS)

    Khan, M.J.H.; Alam, A.B.M.K.; Ahsan, M.H.; Mamun, K.A.A.; Islam, S.M.A.

    2015-01-01

    Highlights: • To validate the reactor physics lattice code WIMSD-5B by this analysis. • To model TRX and BAPL critical experiments using WIMSD-5B. • To compare the calculated results with experiment and MCNP results. • To rely on WIMSD-5B code for TRIGA calculations. - Abstract: The aim of this analysis is to validate the reactor physics lattice transport code WIMSD-5B by TRX (thermal reactor-one region lattice) and BAPL (Bettis Atomic Power Laboratory-one region lattice) critical experiments of light water reactors for neutronics analysis of 3 MW TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh. This analysis is achieved through the analysis of integral parameters of five light water reactor critical experiments TRX-1, TRX-2, BAPL-UO 2 -1, BAPL-UO 2 -2 and BAPL-UO 2 -3 based on evaluated nuclear data libraries JEFF-3.1 and ENDF/B-VII.1. In integral measurements, these experiments are considered as standard benchmark lattices for validating the reactor physics lattice transport code WIMSD-5B as well as evaluated nuclear data libraries. The integral parameters of the said critical experiments are calculated using the reactor physics lattice transport code WIMSD-5B. The calculated integral parameters are compared to the measured values as well as the earlier published MCNP results based on the Chinese evaluated nuclear data library CENDL-3.0 for assessment of deterministic calculation. It was found that the calculated integral parameters give mostly reasonable and globally consistent results with the experiment and the MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are well consistent with each other. Therefore, this analysis reveals the validation study of the reactor physics lattice transport code WIMSD-5B based on JEFF-3.1 and ENDF/B-VII.1 libraries and can also be essential to

  7. Thirty years experience in ET-RR-1 operation

    International Nuclear Information System (INIS)

    Khattab, M.

    1993-01-01

    Egypt research reactor (ET-R R-1) got critical in 27 Th. July 1961. Since that time many technical problems in system manipulation were raised. The present paper demonstrates the famous events that were initiated and the procedures that were taken for their recovery or mitigation. Maintenance problems are discussed. Reactor parameters measuring systems and control have been modernized. The reactor is investigated by in-service inspection program, (ISI). The results show that reactor vessels surface conditions are in good status. experience obtained from these activities and some proposals for future development are summarized. 6 figs., 1 tab

  8. Review of operational experience with the gas-cooled Magnox reactors of the United Kingdom Central Electricity Generating Board

    International Nuclear Information System (INIS)

    Cave, L.; Clarke, A.W.

    1984-01-01

    The paper provides a review, which is mainly of a statistical nature, of 260 reactor years of operating experience which the (United Kingdom) Central Electricity Generating Board (CEGB) has obtained with its gas-cooled, graphite moderated Magnox reactors. The main emphasis in the review is on safety rather than on availability. Data are provided on the overall incidence and frequencies of faults and it is shown that the plant items which are predominantly responsible for recorded faults are the gas circulators and the turbo-alternators. Analysis of the reactor trip experience shows that the incidence of events which necessitate an automatic shutdown of the reactor has been about one per reactor year and that of other events leading to a reactor trip has not been much higher (1.4 per reactor year). As would be expected from the length of the operating experience, some relatively rare events have occurred (expected frequency 10 -2 per reactor year, or less) but on each occasion the reactor shutdown system and decay heat removal systems functioned satisfactorily. No overheating of, or damage to, the fuel occurred as a result of these rare events or of other, more frequent, faults. Analysis of the trend of failure rates has shown an improvement with time in nearly all safety-related items and external inspection of the primary coolant circuits has shown no significant deterioration with time. However, some derating of the reactors has been necessary to reduce the effects of oxidation of mild steel in CO 2 , in order to obtain optimum service lives. In spite of major differences between the systems, a comparison of the failure rates of analogous systems and plant items in PWRs and the Magnox reactors show a considerable similarity. Overall, the review of CEGB's operational experience with its Magnos reactors has shown that the frequencies of faults in systems and plant items has been satisfyingly low. (author)

  9. Design experiences for medical irradiation field at the musashi reactor

    International Nuclear Information System (INIS)

    Aizawa, Otohiko

    1994-01-01

    The design of the medical irradiation field at the Musashi reactor was carried out from 1974 to 1975, about 20 years ago. Various numerical analyses have been carried out recently, and it is astonishing to find out that the performance close to the optimum as a 100 kW reactor has been obtained. The reason for this is that the design was carried out by dividing into the stationary part and the moving part, and as for the moving part, the structure was determined by repeating trial and error and experiments. In this paper, the comparison of the analysis carried out later with the experimental data and the change of the absorbed dose at the time of medical irradiation accompanying the change of neutron energy spectra are reported. As the characteristics of the medical irradiation field at the Musashi reactor, the neutron energy spectra and the absorbed dose and mean medical irradiation time are shown. As the problems in boron neutron capture therapy, the neutron fluence required for the therapy, the way of thinking on background dose, and the problem of determining the irradiation time are discussed. The features of epithermal neutron beam are explained. (K.I.)

  10. Operating experience feedback report: Progress in scram reduction: Commercial power reactors

    International Nuclear Information System (INIS)

    Bell, L.G.; O'Reilly, P.D.

    1989-03-01

    This report documents the results of a trends and patterns analysis of unplanned reactor scrams at commercial US nuclear power reactors from January 1, 1984 to January 1, 1988. Major objectives of this report prepared by the Nuclear Regulatory Commission's (NRC's) Office for Analysis and Evaluation of Operational Data (AEOD) are to: (1) provide feedback of operational experience regarding reactor scram trends in support of the Commission's Strategic Goals, (2) examine the causes of unplanned scrams, and (3) examine the relationship between the causes of unplanned scrams and industry initiatives undertaken to reduce the frequency of unplanned scrams, especially with a view to the potential for future scram rate reduction. 31 refs., 14 figs., 49 tabs

  11. Graphite stack corrosion of BUGEY-1 reactor (synthesis)

    International Nuclear Information System (INIS)

    Petit, A.; Brie, M.

    1996-01-01

    The definitive shutdown date for the BUGEY-1 reactor was May 27th, 1994, after 12.18 full power equivalent years and this document briefly describes some of the feedback of experience from operation of this reactor. The radiolytic corrosion of graphite stack is the major problem for BUGEY-1 reactor, despite the inhibition of the reaction by small quantities of CH 4 added to the coolant gas. The mechanical behaviour of the pile is predicted using the ''INCA'' code (stress calculation), which uses the results of graphite weight loss variation determined using the ''USURE'' code. The weight loss of graphite is determined by annually taking core samples from the channel walls. The results of the last test programme undertaken after the definitive shutdown of BUGEY-1 have enabled an experimental graph to be established showing the evolution of the compression resistance (perpendicular and parallel direction to the extrusion axis) as a function of the weight loss. The numerous analyses, made on the samples carried out in the most sensitive regions, have allowed to verify that no brutal degradation of the mechanical properties of graphite happens for the high value of weight loss up to 40% (maximum weight loss reached locally). (author). 10 refs, 3 figs, 4 tabs

  12. UCN-VCN facility and experiments in Kyoto University Reactor

    International Nuclear Information System (INIS)

    Kawabata, Yuji; Okumura, Kiyoshi; Utsuro, Masahiko

    1993-01-01

    An ultracold and very cold neutron facility was installed in Kyoto University Reactor (KUR). The facility consists of a very cold neutron (VCN) guide tube, a VCN bender, a supermirror neutron turbine and experimental equipments with ultracold neutrons (UCN). The properties of each equipments are presented. UCN is generated by a supermirror neutron turbine combined with the cold neutron source operated with liquid deuterium, and the UCN output spectrum was measured by the time-of-flight method. A gravity analyzer for high resolution spectroscopy and a neutron bottle for decay experiments are now developing as the UCN research in KUR. (author)

  13. MHTGR: New production reactor summary of experience base

    International Nuclear Information System (INIS)

    1988-03-01

    Worldwide interest in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) stems from the capability of the system to retain the advanced fuel and thermal performance while providing unparalleled levels of safety. The small power level of the MHTGR and its passive systems give it a margin of safety not attained by other concepts being developed for power generation. This report covers the experience base for the key nuclear system, components, and processes related to the MHTGR-NPR. 9 refs., 39 figs., 9 tabs

  14. The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow

    Energy Technology Data Exchange (ETDEWEB)

    Vins, M. [Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, V Holesovickach 2, 180 00 Prague 8 (Czech Republic)], E-mail: vinsmiro@seznam.cz

    2008-07-15

    This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)

  15. CANDU fuel - fifteen years of power reactor experience

    International Nuclear Information System (INIS)

    Fanjoy, G.R.; Bain, A.S.

    1977-01-01

    CANDU (Canada Deuterium Uranium) fuel has operated in power reactors since 1962. Analyses of performance statistics, supplemented by examinations of fuel from power reactors and experimental loops have yielded: (a) A thorough understanding of the fundamental behaviour of CANDU fuel. (b) Data showing that the predicted high utilization of uranium has been achieved. Actual fuelling costs in 1976 at the Pickering Generating Station are 1.2 m$/kWh (1976 Canadian dollars) with the simple oncethrough natural-UO 2 fuel cycle. (c) Criteria for operation, which have led to the current very low defect rate of 0.03% of all assemblies and to ''CANLUB'' fuel, which has a graphite interlayer between the fuel and sheath to reduce defects on power increases. (d) Proof that the short length (500 mm), collapsible cladding features of the CANDU bundle are successful and that the fuel can operate at high-power output (current peak outer-element linear power is 58 +- 15% kW/m). Involvement by the utility in all stages of fuel development has resulted in efficient application of this fundamental knowledge to ensure proper fuel specifications, procurement, scheduling into the reactor and feedback to developers, designers and manufacturers. As of mid-1976 over 3 x 10 6 individual elements have been built in a well-estabilished commercially competitive fuel fabrication industry and over 2 x 10 6 elements have been irradiated. Only six defects have been attributed to faulty materials or fabrication, and the use of high-density UO 2 with low-moisture content precluded defects from hydrogen contamination and densification. Development work on UO 2 and other fuel cycles (plutonium and thorium) is continuing, and, because CANDU reactors use on-power fuelling, bundles can be inserted into power reactors for testing. Thus new fuel designs can be quickly adopted to ensure that the CANDU system continues to provide low-cost energy with high reliability

  16. Burnable poisons in the light water reactor design, microburnup experiments and calculations. Part of a coordinated programme on burnup calculations and experiments for thermal reactors

    International Nuclear Information System (INIS)

    Penndorf, K.

    1976-04-01

    Investigations on Research Agreement N 1519/CF (1.8.1974 - 31.7.1975) entitled ''Burnable poisons in light water reactor design, microburnup experiments and calculations'' were carried out in the frame of the IAEA's coordinated research programme on ''Burn-up calculation and experiments for thermal reactors''. The theoretical and experimental work on application of solid burnable poison used for reduction of the amount of boric acid necessary to control of PWR or to lower the number of control rods needed in a BWR. Solid burnable poisons are needed in present PWR designs for the reduction of the boron acid concentration in order to prevent positive coefficients of reactivity. The special operational conditions of a ship reactor lead to the application of this kind of poison for compensation of almost all burnup reactivity. This strengthens the necessity of a very accurate and many dimensional calculations because an appropriate binding of reactivity has to be kept over the whole cycle time. Several burnup experiments had been run in the 15 MW material test reactor FRG-II. The following devices have been irradiated: poison pins within and without PWR fuel pin lattice segments and fuel pins containing pellets with a poison core. Measurements of reactivity, fluence, fission product concentration have been performed. Methods applied were γ-scanning and neutron pulse, radiography and transmission measurement techniques. Evaluation of the experiments was done by one and two dimensional Ssub(N) transport burnup calculations. In parallel a collision probability transport burnup code for current PWR design work is being developed, the main feature of which is economy in manpower and computer time

  17. Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

    International Nuclear Information System (INIS)

    Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.; Powers, Jeffrey J.

    2016-01-01

    Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 "7LiF-BeF_2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

  18. Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Mueller, Donald E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

    2016-08-31

    Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

  19. Vented fuel experiment for gas-cooled fast reactor application

    International Nuclear Information System (INIS)

    Longest, A.W.; Gat, U.; Conlin, J.A.; Campana, R.J.

    1976-01-01

    A pressure-equalized and vented fuel rod is being irradiated in an instrumented capsule designated GB-10 to approximately 100MWd/kg-heavy metal. The fuel is a sol-gel-derived 88 at.% uranium (approximately 9% 235 U) and 12 at.% plutonium oxide, and the cladding is 20% cold-worked 316 stainless steel. The capsule is being irradiated in the Oak Ridge Research Reactor (ORR) and has exceeded a burnup of 70MWd/kg. The fuel has been operated at linear power rates of 39 and 44kW/m, and peak outer cladding temperature of 565 and 630 0 C respectively. A similar fuel rod in a previous capsule (GB-9) was subjected to 48kW/m (685 0 C). Helium gas sweeps through any portion of the three regions of the fuel rod, namely: fuel, blanket, and charcoal trap. The charcoal trap is operated at about 300 0 C. An on-line Ge(Li) detector is used to analyse release rates of several gamma-emitting noble gas isotopes. Analyses are performed primarily on sweep gas flowing through the entire fuel rod, and for sweeps over the top of the charcoal trap. Sweep gas samples are analyzed for stable noble gas isotopes. Results in the form of ratios of release rate over birth rate (R/B) and venting rate over birth rate (V/B) are derived. R/B rates range from 10 -4 % to 30% while V/B ranges from 10 -6 % to 30%. Flow conductance in the capsule was monitored by recording the flow rate and pressure drop across the fuel rod and inlet sweep line. The flow conductance has been falling with increasing burnup, currently restricting the flow to about 20ml (s.t.p.)/min at a pressure difference of about 1.5MPa. Venting rates of the gaseous fission products as a function of gas pressure in the range 6.9 to 1.4MPa have also been measured. Planned future experiments include the monitoring of tritium release, venting and cladding permeation rates, and its molecular form. First measurements have been made. A simulated leak experiment will determine the mixture of fission gases as a function of flow rate and the most

  20. Maximizing the use of research reactors in training power reactor operating staff with special reference to US experience

    International Nuclear Information System (INIS)

    Cox, J.A.

    1976-01-01

    Research reactors have been used in training nuclear power plant personnel for many years. Using the experience in the United States of America a programme is proposed that will maximize the training conducted at a research reactor and lessen the time that the staff must spend training elsewhere. The programme is adaptable to future training of replacement staff and for staff retraining. (author)

  1. Neutronics analysis of Nigerian Research Reactor-1

    International Nuclear Information System (INIS)

    Azande, T.S.; Balogun, G.I.

    2010-01-01

    Feasibility studies for the conversion of the Nigerian Research Reactor-1 (NIRR-1) have been performed using WIMS and CITATION codes (Azande et al, 2009 and Balogun, 2003) at the Centre for Energy Research and Training (CERT), Ahmadu Bello University, Zaria Kaduna State. In this work, the neutronics analysis of NIRR-1 core concerning mass loading of U-235 in the core, shut down margin (SDM), safety reactivity factor (SRF), control rod worth, and control rod critical depth of insertion were investigated at low enrichment. Two fuel types (UAl 4 and UO 2 ) were considered and the uranium densities required for the conversion of NIRR-1 core to low enrichment were computed to be 1201g/cc with 20% enrichment, 1144 g/cc with 19.75% enrichment, 1274 g/cc with 15% enrichment, 1448 g/cc with 10% enrichment for UAl 4 fuel type and 1141g/cc with 20% enrichment, 1144 g/cc with 19.75% enrichment, 1216 g/cc with 15% enrichment, and 1389 g/cc with 10% enrichment for UO 2 fuel type. Signi ficantly, higher uranium densities are required to convert NIRR-1 from HEU to LEU - indicating a drastic review of the NIRR-1 core.

  2. Thai Research Reactor (TRR-1/M1) Neutron Beam Measurements

    International Nuclear Information System (INIS)

    Ratanatongchai, Wichian

    2009-07-01

    Full text: Neutron beam tube of neutron radiography facility at Thai Research Reactor (TRR-1/M1) Thailand Institute of Nuclear Technology (public organization) is a divergent beam. The rectangular open-end of the beam tube is 16 cm x 17 cm while the inner-end is closed to the reactor core. The neutron beam size was measured using 20 cm x 40 cm neutron imaging plate. The measurement at the position 100 cm from the end of the collimator has shown that the beam size was 18.2 cm x 19.0 cm. Gamma ray in neutron the beam was also measured by the identical position using industrial X ray film. The area of gamma ray was 27.8 cm x 31.1 cm with the highest intensity found to be along the neutron beam circumference

  3. Problems of nuclear reactor safety. Vol. 1

    International Nuclear Information System (INIS)

    Shal'nov, A.V.

    1995-01-01

    Proceedings of the 9. Topical Meeting 'Problems of nuclear reactor safety' are presented. Papers include results of studies and developments associated with methods of calculation and complex computerized simulation for stationary and transient processes in nuclear power plants. Main problems of reactor safety are discussed as well as rector accidents on operating NPP's are analyzed

  4. Moessbauer experiments, ch. 1

    International Nuclear Information System (INIS)

    Maring, K.W.

    1977-01-01

    The magnetic properties of iron base alloys have been investigated as a function of the concentrations of the sp-elements Al, Si, Ga, Ge, As, Sn, and Sb. An extensive description of the Moessbauer Effect Spectroscopy (MES) technique is given. The results of MES experiments at room temperature for FeX alloys (X stands for one of the sp-elements) and impurity concentrations of 1, 3 and 5 a/o are presented. Through temperature dependant measurements the hyperfine fields H ((m,n)T) at iron nuclei in different surroundings are obtained. The change of the Curie temperature of the various dilute iron-base alloys, with respect to pure iron, as a function of the impurity concentration is investigated. Average magnetic hyperfine fields and isomer shifts measured at 119 Sn nuclei in Fe 119 Sn are discussed

  5. The combined use of test reactor experiments and power reactor tests for the development of PCI-resistant fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Vesterlund, G.; Vaernild, O.

    1980-01-01

    The theme of this paper is that for development of PCI-resistant fuel acceptable from the commercial and licensing aspects, extensive and time-consuming work is needed both in a test reactor and in power reactors. The test reactor is necessary for ramp testing to power levels not allowed in power reactors and with the aim of generating fuel failures. It is also used for other special irradiation experiments. The access to power reactors is necessary to generate information on performance in a real LWR core and to incubate at a reasonable cost the large amount of rods required for test reactor ramping. Selected results from the ASEA-ATOM work are used to support these conclusions. (author)

  6. International feedback experience on the cutting of reactor internal components

    International Nuclear Information System (INIS)

    Boucau, J.

    2014-01-01

    Westinghouse capitalizes more than 30 years of experience in the cutting of internal components of reactor and their packaging in view of their storage. Westinghouse has developed and validated different methods for cutting: plasma torch cutting, high pressure abrasive water jet cutting, electric discharge cutting and mechanical cutting. A long feedback experience has enabled Westinghouse to list the pros and cons of each cutting technology. The plasma torch cutting is fast but rises dosimetry concerns linked to the control of the cuttings and the clarity of water. Abrasive water jet cutting requires the installation of costly safety devices and of an equipment for filtering water but this technology allows accurate cuttings in hard-to-reach zones. Mechanical cutting is the most favourable technology in terms of wastes generation and of the clarity of water but the cutting speed is low. (A.C.)

  7. Chemistry in water reactors: operating experience and new developments. 2 volumes

    International Nuclear Information System (INIS)

    1994-01-01

    These proceedings of the International conference on chemistry in water reactors (Operating experience and new developments), Volume 1, are divided into 8 sessions bearing on: (session 1) Primary coolant activity, corrosion products (5 conferences), (session 2) Dose reduction (4 conferences), (session 3) New developments (4 conferences), poster session: Primary coolant chemistry (16 posters), (session 4) Decontamination (5 conferences), poster session (2 posters), (session 5) BWR-Operating experience (3 conferences), (session 6) BWR-Modelling of operating experience (4 conferences), (session 7) BWR-Basic studies (4 conferences), (session 8) BWR-New technologies (3 conferences)

  8. Status of prompt gamma neutron activation analysis (PGAA) at TRR-1/M1 (Thai Research Reactor-1/Modified 1)

    International Nuclear Information System (INIS)

    Asvavijnijkulchai, Chanchai; Dharmavanij, Wanchai; Siangsanan, Pariwat; Ratanathongchai, Wichian; Chongkum, Somporn

    1999-01-01

    The first prompt gamma activation analysis (PGAA) was designed, constructed and installed at a 6 inch diameter neutron beam port of the Thai Research Reactor-1/Modified 1 (TRR-1/M1) since 1989. Beam characteristic were made by Gd foil irradiation, X-ray film exposing and densitometry scanning consequently. The thermal neutron flux at sample position was measured by Au foil activation, and was about 1 x 10 7 n.cm 2 .sec -1 at 700 kW operating power. The experiments have been conducted successfully. In 1998, the PGAA facility has been developed for the reactor operating power at 1.2 MW. The new PGAA system, e.g., beam shutter, gamma collimator and biological shields have been designed to reduce the leakage of neutrons and gamma radiation to the acceptance levels in accordance with the International Commission on Radiation Protection Publication 60 (ICRP 60). The construction and installation will be completed in April 1999. (author)

  9. Management of historical waste from research reactors: the Dutch experience

    International Nuclear Information System (INIS)

    Van Heek, Aliki; Metz, Bert; Janssen, Bas; Groothuis, Ron

    2013-01-01

    Most radioactive waste emerges as well-defined waste streams from operating power reactors. The management of this is an on-going practice, based on comprehensive (IAEA) guidelines. A special waste category however consists of the historical waste from research reactors, mostly originating from various experiments in the early years of the nuclear era. Removal of the waste from the research site, often required by law, raises challenges: the waste packages must fulfill the acceptance criteria from the receiving storage site as well as the criteria for nuclear transports. Often the aged waste containers do not fulfill today's requirements anymore, and their contents are not well documented. Therefore removal of historical waste requires advanced characterization, sorting, sustainable repackaging and sometimes conditioning of the waste. This paper describes the Dutch experience of a historical waste removal campaign from the Petten High Flux research reactor. The reactor is still in operation, but Dutch legislation asks for central storage of all radioactive waste at the COVRA site in Vlissingen since the availability of the high- and intermediate-level waste storage facility HABOG in 2004. In order to comply with COVRA's acceptance criteria, the complex and mixed inventory of intermediate and low level waste must be characterized and conditioned, identifying the relevant nuclides and their activities. Sorting and segregation of the waste in a Hot Cell offers the possibility to reduce the environmental footprint of the historical waste, by repackaging it into different classes of intermediate and low level waste. In this way, most of the waste volume can be separated into lower level categories not needing to be stored in the HABOG, but in the less demanding LOG facility for low-level waste instead. The characterization and sorting is done on the basis of a combination of gamma scanning with high energy resolution of the closed waste canister and low

  10. Experience in using a research reactor for the training of power reactor operators

    International Nuclear Information System (INIS)

    Blotcky, A.J.; Arsenaut, L.J.

    1972-01-01

    A research reactor facility such as the one at the Omaha Veterans Administration Hospital would have much to offer in the way of training reactor operators. Although most of the candidates for the course had either received previous training in the Westinghouse Reactor Operator Training Program, had operated nuclear submarine reactors or had operated power reactors, they were not offered the opportunity to perform the extensive manipulations of a reactor that a small research facility will allow. In addition the AEC recommends 10 research reactor startups per student as a prerequisite for a cold operator?s license and these can easily be obtained during the training period

  11. IRPhE - International Reactor Physics Experiments database

    International Nuclear Information System (INIS)

    Sartori, E.

    2004-01-01

    The OECD/NEA Nuclear Science Committee (NSC) has identified the need to establish international databases containing all the important experiments that are available for sharing among the specialists and has set up or sponsored specific activities to achieve this. The aim is to preserve them in an agreed standard format in computer accessible form, to use them for international activities involving validation of current and new calculational schemes including computer codes and nuclear data libraries, for assessing uncertainties, confidence bounds and safety margins, and to record measurement methods and techniques. It is a significant saving results from disseminating a standard benchmark set to be used worldwide. A framework for professionals that use the standard benchmark set to validate and verify modeling codes and data for radiation transport, criticality safety and reactor physics applications guarantees a comparative set of analyses. It represents also a good basis for pinpointing important gaps and where efforts should be concentrated and ensures knowledge and competence preservation, management and transfer in nuclear science and engineering. A large number of experimentalists, physicists, evaluators, modelers have devoted large amounts of their efforts and competencies to produce the data on which the methods we are using today are based. These data are far from having been exploited fully for the different nuclear and radiation technologies. This wealth of information needs to be preserved in a form more easily exploitable by modern information technology and for use in connection with novel and refined computational models with limitations of the past removed. These data will form the basis for the studies of more advanced nuclear technology, will be instrumental in identifying areas where there is a lack of knowledge and thus provide support to justifying new experiments that would reduce design uncertainties and consequently costs. Improvement of

  12. Stresses in reactor pressure vessel nozzles -- Calculations and experiments

    International Nuclear Information System (INIS)

    Brumovsky, M.; Polachova, H.

    1995-01-01

    Reactor pressure vessel nozzles are characterized by a high stress concentration which is critical in their low-cycle fatigue assessment. Program of experimental verification of stress/strain field distribution during elastic-plastic loading of a reactor pressure vessel WWER-1000 primary nozzle model in scale 1:3 is presented. While primary nozzle has an ID equal to 850 mm, the model nozzle has ID equal to 280 mm, and was made from 15Kh2NMFA type of steel. Calculation using analytical methods was performed. Comparison of results using different analytical methods -- Neuber's, Hardrath-Ohman's as well as equivalent energy ones, used in different reactor Codes -- is shown. Experimental verification was carried out on model nozzles loaded statically as well as by repeated loading, both in elastic-plastic region. Strain fields were measured using high-strain gauges, which were located in different distances from center of nozzle radius, thus different stress concentration values were reached. Comparison of calculated and experimental data are shown and compared

  13. Ageing problems and renovation programme of ET-RR-1 reactor

    International Nuclear Information System (INIS)

    Khattab, M.S.; Sultan, M.A.

    1995-01-01

    Based on Practical Experience gained from interfacing ageing systems in addition to operating new systems, current problems could be deduced whenever in-service inspection are carried out. This paper summarizes the in-service inspection made, and the proposed programme of rehabilitation of mechanical system in the ET-RR-1 research reactor at Inshass. Exchangeable experience in solving common problems in similar reactors play an important role in the effectiveness of such rehabilitation programme. The paper summarizes also the modernization of control, measuring and radiation monitoring system already carried out at the reactor. (orig.)

  14. Extensive utilisation of VR-1 reactor for nuclear education and training

    International Nuclear Information System (INIS)

    Rataj, J.

    2010-01-01

    The paper presents utilisation of the VR-1 reactor for nuclear education and training at national and international level. VR-1 reactor has been operating by the Czech Technical University since December 1990. The reactor is a pool-type light water reactor based on enriched uranium (19.7% 235 U) with maximum thermal power 1kW and for short time period up to 5kW. The moderator of neutrons is light water, which is also used as a reflector, a biological shielding and a coolant. Heat is removed from the core by natural convection. The pool disposition of the reactor facilitates access to the core, setting and removing of various experimental samples and detectors, easy and safe handling of fuel assemblies. The reactor core can contain from 17 to 21 fuel assemblies IRT-4M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The reactor is equipped with several experimental devices; e.g. horizontal, radial and tangential channels used to take out a neutron beam, reactivity oscillator for dynamics study and bubble boiling simulator. The reactor has been used very efficiently especially for education and training of university students and NPP's specialists for more than 18 years. The VR-1 reactor is utilised within various national and international activities such as Czech Nuclear Education Network (CENEN), European Nuclear Education Network and also Eastern European Research Reactor Initiative (EERRI). The reactor is well equipped for education and training not only by the experimental facility itself but also by incessant development of training methods and improvement of education experiments. The education experiments can be combined into training courses attended by students according to their study specialization and knowledge level. The training programme is aimed to the reactor and neutron physics, dosimetry, nuclear safety, and control of nuclear installations. Every year, approximately 250 university students undergo

  15. NRC data base for power reactor surveillance programs and for irradiation experiments results

    International Nuclear Information System (INIS)

    Kam, F.B.K.; Stallmann, F.W.

    1991-01-01

    The radiation damage of pressure vessel materials in nuclear reactors depends on many different factors, primarily fluence, fluence spectrum, fluence rate, irradiation temperature, and chemistry. These factors and, possibly, others such as heat treatment and type of flux used in weldments must be considered to reliably predict the pressure vessel embrittlement and to assure the safe operation of the reactor. Based on embrittlement predictions, decisions must be made concerning operating parameters, low-leakage fuel management, possible life extension, and the need for annealing of the pressure vessel. Large numbers of data obtained from surveillance capsules and test reactor experiments are needed, comprising many different materials and different irradiation conditions, to develop generally applicable damage prediction models that can be used for industry standards and regulatory guides. The US Nuclear Regulatory Agency has, therefore, sponsored a project to construct an Embrittlement Data Base (EDB) for a comprehensive collection of data concerning changes in material properties of pressure vessel steels due to neutron irradiation. A first version containing data from surveillance capsules of commercial power reactors, the Power Reactor Embrittlement Data Base (PR-EDB) Version 1, has been completed and is available to authorized users from the Radiation Shielding Information Center at the Oak Ridge National Laboratory. This document provides a discussion of the features of the current database. 1 fig

  16. Reactor Engineering Department annual report, April 1, 1985 - March 31, 1986

    International Nuclear Information System (INIS)

    1986-08-01

    Research and development activities in the Department of Reactor Engineering in fiscal 1985 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor, High Conversion Light Water Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, reactor decommissioning technology, and activities of the Committee on Reactor Physics. (author)

  17. Advanced Reactor Licensing: Experience with Digital I&C Technology in Evolutionary Plants

    Energy Technology Data Exchange (ETDEWEB)

    Wood, RT

    2004-09-27

    This report presents the findings from a study of experience with digital instrumentation and controls (I&C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l&C systems and identified lessons learned. The report (1) gives an overview of the modern l&C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States.

  18. Advanced Reactor Licensing: Experience with Digital I and C Technology in Evolutionary Plants

    International Nuclear Information System (INIS)

    Wood, RT

    2004-01-01

    This report presents the findings from a study of experience with digital instrumentation and controls (I and C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l and C systems and identified lessons learned. The report (1) gives an overview of the modern l and C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States

  19. Experiment for search for sterile neutrino at SM-3 reactor

    Science.gov (United States)

    Serebrov, A. P.; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Cherniy, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Zinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanasiev, V. V.; Matrosov, L. N.; Matrosova, M. Yu.

    2016-11-01

    In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

  20. International Reactor Physics Experiment Evaluation (IRPhE) Project. IRPhE Handbook - 2017 edition

    International Nuclear Information System (INIS)

    2017-01-01

    The International Reactor Physics Evaluation (IRPhE) Project was initiated as a pilot in 1999 by the Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June 2003. While the NEA co-ordinates and administers the IRPhE Project at the international level, each participating country is responsible for the administration, technical direction and priorities of the project within their respective countries. The information and data included in this handbook are available to NEA member countries, to all contributing countries and to others on a case-by-case basis. The IRPhE Project is patterned after the International Criticality Safety Benchmark Evaluation Project (ICSBEP). It closely co-ordinates with the ICSBEP to avoid duplication of efforts and publication of conflicting information. Some benchmark data are applicable to both nuclear criticality safety and reactor physics technology. Some have already been evaluated and published by the ICSBEP, but have been extended to include other types of measurements in addition to the critical configuration. Through this effort, the IRPhE Project will be able to 1) consolidate and preserve the existing worldwide information base; 2) retrieve lost data; 3) identify areas where more data are needed; 4) draw upon the resources of the international reactor physics community to help fill knowledge gaps; 5) identify discrepancies between calculations and experiments due to deficiencies in reported experimental data, cross-section data, cross-section processing codes and neutronics codes; 6) eliminate a large amount of redundant research and processing of reactor physics experiment data, and 7) improve future experimental planning, execution and reporting. This handbook contains reactor physics benchmark specifications that have been derived from experiments performed at nuclear facilities around the world. The benchmark specifications are intended for use by

  1. Irradiation Experiments on Plutonium Fuels for Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Frost, B. R.T.; Wait, E. [Atomic Energy Research Establishment Harwell, Berks. (United Kingdom)

    1967-09-15

    An assessment carried out some years ago indicated that cermet fuels might provide the high burn-up and integrity required for fast reactors. An irradiation programme was started at Harwell on (U, Pu)O{sub 2} -SS cermet plates and rods, mainly In thermal neutron fluxes, to gain experience of dimensional stability at temperatures typical of modern sodium-cooled fast reactor designs (600-650 Degree-Sign C). A subsequent assessment showed that cermets carried a large penalty as far as breeding was concerned and (U, Pu)C was chosen by Harwell for long-term study as an alternative, economic, fast reactor fuel. However, the results from the cermet experiments were of sufficient promise to proceed with parallel irradiation programmes on cermets and carbide. The studies of cermets showed that dimensional instability (swelling and cladding rupture) were caused by the pressures exerted on the steel matrix by the fuel particles, and that the initial density of the fuel particles was important in determining the burn-up at which failure occurred. Further, it was shown that cermets provided a useful vehicle for studying the changes occurring in oxide fuel particles with increasing burn-up. The disappearance of initial porosity and its replacement by fission gas bubbles and segregated solid fission products was studied in some detaiL No significant differences were observed between UO{sub 2} and(U,Pu)O{sub 2} particles. The initial studies of (U, Pu)C were concerned with the effect of varying composition and structure on swelling and fission gas release. A tantalum-lined nickel alloy cladding material was used to contain both pellet and powder specimens In an irradiation experiment in the core of the Dounreay fast reactor. This showed that the presence of a metal phase in the fuel led to a high swelling rate, that fission gas release was low up to {approx} 3% bum-up, and that a low density powder accommodated the swelling without excessive straining of the can. A subsequent

  2. The Program Planned for the Molten Salt Reactor Experiment

    International Nuclear Information System (INIS)

    Haubenreich, Paul N.

    1967-01-01

    This document outlines the program planned for the MSRE in fiscal years 1968 and 1969. It includes a bar diagram of the program, a critical-path type diagram of the operations, and a brief description of each task. In addition to the work at the reactor site, the outline also covers activities elsewhere at ORNL and by the AEC that directly affect the reactor schedule. The amount of detail and the accuracy with which we can estimate times varies considerably among the different items on the schedule. Some items, such as annual checkouts and core sample replacement, have been done before and our time estimates do not include any contingency, In the case of such tasks as planning, reviewing, and preparing for experiments or operations, we have set target dates that appear reasonable and we fully expect to meet these. Processing the salt is a different matter. If there are no unforeseen difficulties we should finish easily in the time shown, but the operation is in part a shakedown, so delays would not be too surprising, The time for modifying the system and adding fluoroborate is, of course, uncertain because the requirements are not yet known. As the requirements develop in more detail the estimate will be updated, but we do not foresee any major dislocation in the schedule, The scheduled time for preparation of enriching salt is becoming tight because of delays in facility construction. Should there be further delays in this key item, the entire schedule would have to be reconsidered.

  3. Experiment calculated ascertainment of factors affecting the energy release in IGR reactor core

    International Nuclear Information System (INIS)

    Kurpesheva, A.M.; Zhotabayev, Zh.R.

    2006-01-01

    Full text: At present energy supply resources problem is important. Nuclear reactors can, of course, solve this problem, but at the same time there is another issue, concerning safety exploitation of nuclear reactors. That is why, for the last seven years, such experiments as 'Investigation of the processes, conducting severe accidents with core melting' are being carried out at our IGR (impulse graphite reactor) reactor. Leaving out other difficulties of such experiments, it is necessary to notice, that such experiments require more accurate IGR core energy release calculations. The final aim of the present research is verification and correction of the existing method or creation of new method of IGR core energy release calculation. IGR reactor is unique and there is no the same reactor in the world. Therefore, application of the other research reactor methods here is quite useful. This work is based on evaluation of factors affecting core energy release (physical weight of experimental device, different configuration of reactor core, i.e. location of absorbers, initial temperature of core, etc), as well as interference of absorbers group. As it is known, energy release is a value of integral reactor power. During experiments with rays, Reactor power depends on currents of ion production chambers (IPC), located round the core. It is worth to notice that each ion production chamber (IPC) in the same start-up has its own ratio coefficient between IPC current and reactor present power. This task is complicated due to 'IPC current - reactor power' ratio coefficients, that change continuously, probably, because of new loading of experimental facility and different position of control rods. That is why, in order to try about reactor power, before every start-up, we have to re-determine the 'IPC current - reactor power' ratio coefficients for each ion production chamber (IPC). Therefore, the present work will investigate the behavior of ratio coefficient within the

  4. Regulatory aspects and experience with Russian research reactors

    International Nuclear Information System (INIS)

    Morozov, S.I.

    2003-01-01

    Regulatory activity of Gosatomnadzor of Russia in the field of research reactors (RR) safety implies implementing three major aspects: 1) establishing the nuclear and radiation safety standards; 2) licensing; and 3) inspection and enforcement. Relatively recently a full set of safety standards and regulations for RR has been established thus allowing Gosatomnadzor of Russia to effectively implement its designated functions in the field of RR safety. A minimum set of these documents is shown as follows: Level I: Fundamentals: Law 'On the use of nuclear energy'; Law 'On Public radiation protection' Level II: Safety Standard: 'General Provisions for Safety of Research Facilities' Level III: Safety Rules: Nuclear Safety; - Radiation Safety; Waste Management; Safe Decommissioning of RR; Safety Analysis Report; QAP Level IV: Safety Regulations: Licensing (incl. Peer Review and Safety Assessment) - Inspection Gosatomnadzor of Russia has created and regularly updates the database on nuclear research reactors based on the actual status of all facilities. According to the database many facilities have been shutdown during recent years whether temporary or permanently waiting for the final decision on their decommissioning. For example, in 2003 Gosatomnadzor of Russia has 85 nuclear research reactors under its supervision (compared to 113 in 1998). This fact can be explained by three main reasons: 1) experimental program finished and no other programmes in place; 2) lack of resources (financial and human); 3) safety problems (physical obsolescence and ageing of equipment). One of the main difficulties in regulating RR safety is a variety of operating organizations - 21, with different financial and human resource capabilities. Ministries responsible for supporting their operation are of a little help. It becomes obvious that a unified governmental program for RR utilization is urgently needed to decide what number of RR and for what needed purposes is required to support the

  5. Experiments utilizing two coupled TRIGA-type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Thayer, G [Southern California Edison Co., Rosemead, CA (United States); Jones, B G; Miley, G H [University of Illinois (United States)

    1974-07-01

    An experimental study has been performed on a coupled-core system consisting of two reactors each of which can be made critical by itself, coupled neutronically by a graphite thermal column. Both steady-state and transient measurements were performed on the system. The steady-state measurement consisted of measuring the coupling coefficient between the two reactors. Also, series of measurements were performed while one of the cores was far subcritical and the coupling between the two cores was varied between 1.6 x 10{sup -2} and 1.6 x 10{sup -5} cents by the insertion of a water gap and from 1.6 x 10{sup -2} cents to 6.0 x 10{sup -4} cents by the insertion of a cadmium sheet between the cores. The transient portion of the study was performed by pulsing one of the reactors (the Illinois Advanced TRIGA) and following the pulse into the passive core (the Low Power Reactor Assembly). The first pulse series measured the pulse as it emerged from the thermal column and propagated through the water, where no fuel was present. This provided an analysis of the neutron source to the passive core. The second pulse series was performed with the passive core far subcritical (k{sub eff} {approx_equal} 0.94) and investigated the effects on the transient coupling of the insertion of water gaps of up to 9 inches or a cadmium sheet ({sigma}T = 3.2) between the two cores. Spatial measurements of the pulse in the far subcritical assembly also were performed. The third series of pulses investigated the characteristics of the pulse in the passive core when it was subcritical, just critical, and supercritical, The effects on the FWHM of the pulse in the passive core and on the delay time between the peak of the pulse in the TRIGA and the passive core were measured for the passive core having a k{sub eff} from 0.936 to 1.0015 and the initial period of the pulse in TRIGA varying from 15.6 {+-} .7 ms to 3.58 {+-} .05 ms. The FWHM increased from 13.5 {+-} 0.5 ms to 18.8 {+-} 0.5 ms and delay

  6. Experience of TTR-1 decommissioning

    International Nuclear Information System (INIS)

    Kato, Hiroaki; Nobuoka, Yoshishige; Yoshimura, Yukio; Homma, Hitoshi; Nakai, Masaru

    2005-01-01

    Toshiba Training Reactor-1 (TTR-1) was planned for improvement of technical level from the standpoint of nuclear reactor manufacturer, training of a nuclear engineer, and research of nuclear physics, radiochemistry, radiation shielding and others. TTR-1 was permitted for construction in May 1960, attained at the first criticality in March 1962 and has continued to operate over 40 years. TTR-1 was permanently shut down in March 2001, accomplishing the planned target. From the initial criticality to the shut down, total operating time amounts to 15,300 hours and 31 MWds. Decommissioning plan was submitted to the Ministry of Education, Culture, Sports, Science and Technology on August 8, 2001 and dismantling work was started. The spent fuel was transported outside the laboratory, and the first phase and the second phase dismantling work were completed at the end of February 2004. Some of the reactor equipments continue maintaining their performance, and waste materials generated from dismantling work are under the state of managed storage, until disposal of the dismantling radioactive waste becomes clear, when the third phase of dismantling work will be started. At the end of the third phase work, all the TTR-1 equipments are dismantled and all waste materials are removed from TTR-1, then decommissioning of TTR-1 is completed. The outline of the decommissioning plan, the actually performed dismantling work, and spent fuel transportation work is briefly described. (author)

  7. Capacitor requirements for controlled thermonuclear experiments and reactors

    International Nuclear Information System (INIS)

    Boicourt, G.P.; Hoffman, P.S.

    1975-01-01

    Future controlled thermonuclear experiments as well as controlled thermonuclear reactors will require substantial numbers of capacitors. The demands on these units are likely to be quite severe and quite different from the normal demands placed on either present energy storage capacitors or present power factor correction capacitors. It is unlikely that these two types will suffice for all necessary Controlled Thermonuclear Research (CTR) applications. The types of capacitors required for the various CTR operating conditions are enumerated. Factors that influence the life, cost and operating abilities of these types of capacitors are discussed. The problems of capacitors in a radiation environment are considered. Areas are defined where future research is needed. Some directions that this research should take are suggested. (U.S.)

  8. Reactor Dynamics Experiments with a Sub-Critical Assembly

    International Nuclear Information System (INIS)

    Miley, G.H.; Yang, Y.; Wu, L.; Momota, H.

    2004-01-01

    A resurgence in use of nuclear power is now underway worldwide. However due to the shutdown of many university research reactors , student laboratories must rely more heavily on use of sub-critical assemblies. Here a driven sub-critical is described that uses a cylindrical Inertial Electrostatic Confinement (IEC) device to provide a fusion neutron source. The small IEC neutron source would be inserted in a fuel element position, with its power input controlled externally at a control panel. This feature opens the way to use of the critical assembly for a number of transient experiments such as sub-critical pulsing and neutron wave propagation. That in turn adds important new insights and excitement for the student teaching laboratory

  9. Experiments on Critical Heat Flux for CAREM -25 Reactor

    International Nuclear Information System (INIS)

    Mazufri, C.M

    2000-01-01

    The prediction of critical heat flux (CHF) in rod bundles of light water reactors is basically performed with the aid of empirical correlations derived from experimental data.Many CHF correlations have been proposed and are widely used in the analysis of the thermal margin during normal operation, transient, and accident conditions.Correlations found in the open literature are not sufficiently verified for the thermal hydraulic conditions that appear in the CAREM core under normal operation: high pressure, low flow, and low qualities.To compensate this deficiency, an experimental investigation on CHF in such thermal-hydraulic conditions was carried out.The experiments have been performed in the Institute of Physics and Power Engineering of Russian Federation.A short description of facilities, details of the experimental program and some preliminary results obtained are presented in this work

  10. Capacitor requirements for controlled thermonuclear experiments and reactors

    International Nuclear Information System (INIS)

    Boicourt, G.P.; Hoffman, P.S.

    1975-01-01

    Future controlled thermonuclear experiments as well as controlled thermonuclear reactors will require substantial numbers of capacitors. The demands on these units are likely to be quite severe and quite different from the normal demands placed on either present energy storage capacitors or present power factor correction capacitors. It is unlikely that these two types will suffice for all necessary Controlled Thermonuclear Research (CTR) applications. The types of capacitors required for the various CTR operating conditions are enumerated. Factors that influence the life, cost and operating abilities of these types of capacitors are discussed. The problems of capacitors in a radiation environment are considered. Areas are defined where future research is needed. Some directions that this research should take are suggested

  11. Reactor Emergency Action Level Monitor: Knowledge acquisition experiences

    International Nuclear Information System (INIS)

    Touchton, R.A.

    1987-01-01

    This paper presents the Knowledge Acquisition experiences in developing the Reactor Emergency Action Level Monitor (REALM) Expert System Prototype. REALM is an expert system which interprets plant sensor data and provides advice on the proper emergency classification. The REALM project is being funded by the Electric Power Research Institute, Consolidated Edison is serving as the host utility, and the effort is being conducted by Technology Applications, Inc. REALM is being designed to provide expert assistance in the identification of a nuclear power plant emergency situation and the determination of its severity, ultimately operating in a real-time, on-line processing environment. The paper discusses briefly the direct knowledge acquisition techniques used by the project team (who are themselves power industry engineers), to extract relevant knowledge from plant specifications and procedures

  12. Analysis of kyoto university reactor physics critical experiments using NCNSRC calculation methodology

    International Nuclear Information System (INIS)

    Amin, E.; Hathout, A.M.; Shouman, S.

    1997-01-01

    The kyoto university reactor physics experiments on the university critical assembly is used to benchmark validate the NCNSRC calculations methodology. This methodology has two lines, diffusion and Monte Carlo. The diffusion line includes the codes WIMSD4 for cell calculations and the two dimensional diffusion code DIXY2 for core calculations. The transport line uses the MULTIKENO-Code vax Version. Analysis is performed for the criticality, and the temperature coefficients of reactivity (TCR) for the light water moderated and reflected cores, of the different cores utilized in the experiments. The results of both Eigen value and TCR approximately reproduced the experimental and theoretical Kyoto results. However, some conclusions are drawn about the adequacy of the standard wimsd4 library. This paper is an extension of the NCNSRC efforts to assess and validate computer tools and methods for both Et-R R-1 and Et-MMpr-2 research reactors. 7 figs., 1 tab

  13. Experimental study of the temperature distribution in the TRIGA IPR-R1 Brazilian research reactor

    International Nuclear Information System (INIS)

    Mesquita, Amir Zacarias

    2005-01-01

    The TRIGA-IPR-R1 Research Nuclear Reactor has completed 44 years in operation in November 2004. Its initial nominal thermal power was 30 kW. In 1979 its power was increased to 100 kW by adding new fuel elements to the reactor. Recently some more fuel elements were added to the core increasing the power to 250 kW. The TRIGA-IPR-R1 is a pool type reactor with a natural circulation core cooling system. Although the large number of experiments had been carried out with this reactor, mainly on neutron activation analysis, there is not many data on its thermal-hydraulics processes, whether experimental or theoretical. So a number of experiments were carried out with the measurement of the temperature inside the fuel element, in the reactor core and along the reactor pool. During these experiments the reactor was set in many different power levels. These experiments are part of the CDTN/CNEN research program, and have the main objective of commissioning the TRIGA-IPR-R1 reactor for routine operation at 250 kW. This work presents the experimental and theoretical analyses to determine the temperature distribution in the reactor. A methodology for the calibration and monitoring the reactor thermal power was also developed. This methodology allowed adding others power measuring channels to the reactor by using thermal processes. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were also experimentally valued. lt was also presented a correlation for the gap conductance between the fuel and the cladding. The experimental results were compared with theoretical calculations and with data obtained from technical literature. A data acquisition and processing system and a software were developed to help the investigation. This system allows on line monitoring and registration of the main reactor operational parameters. The experiments have given better comprehension of the reactor thermal-fluid dynamics and helped to develop numerical

  14. The double chooz experiment: simulation of reactor antineutrino spectra

    International Nuclear Information System (INIS)

    Mueller, T.

    2010-01-01

    The Double Chooz experiment aims to study the oscillations of electron antineutrinos produced by the Chooz nuclear power station, located in France, in the Ardennes region. It will lead to an unprecedented accuracy on the value of the mixing angle θ 13 . Improving the current knowledge on this parameter, given by the Chooz experiment, requires a reduction of both statistical and systematic errors, that is to say not only observing a large data sample, but also controlling the experimental uncertainties involved in the production and detection of electron antineutrinos. The use of two identical detectors will cancel most of the experimental systematic uncertainties limiting the sensitivity to the value of the mixing angle. We present in this thesis, simulations of reactor antineutrino spectra that were carried out in order to control the sources of systematic uncertainty related to the production of these particles by the plant. We also discuss our work on controlling the normalization error of the experiment through the precise determination of the number of target protons by a weighing measurement and through the study of the fiducial volume of the detectors which requires an accurate modeling of neutron physics. After three years of data taking with two detectors, Double Chooz will be able to disentangle an oscillation signal for sin 2 2θ 13 ≥ 0.05 (at 3σ) or, if no oscillations are observed, to put a limit of sin 2 2θ 13 ≤ 0.03 at 90% C.L. (author) [fr

  15. Upgrading of the research reactors FRG-1 and FRG-2

    International Nuclear Information System (INIS)

    Krull, W.

    1981-01-01

    In 1972 for the research reactor FRG-2 we applied for a license to increase the power from 15 MW to 21 MW. During this procedure a public laying out of the safety report and an upgrading procedure for both research reactors - FRG-1 (5 MW) and FRG-2 - were required by the licensing authorities. After discussing the legal background for licensing procedures in the Federal Republic of Germany the upgrading for both research reactors is described. The present status and future licensing aspects for changes of our research reactors are discussed, too. (orig.) [de

  16. Annual report on JEN-1 and JEN-2 Reactors

    International Nuclear Information System (INIS)

    Montes Ponce de Leon, J.

    1974-01-01

    In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

  17. Experiment WA1 (CDHS Neutrino Experiment)

    CERN Multimedia

    CERN PhotoLab

    1977-01-01

    Experiment WA1, also known under CDHS (CERN, Dortmund, Heidelberg, Saclay; spokesman Jack Steinberger), was the first neutrino experiment on the SPS, in its West Area. Magnetized iron (with a toroidal field) forms the core of the detector. On its outside we see drift chambers and photomultipliers (detecting the light from the plastic scintillators further in). Peter Schilly is wearing a white coat. See also CERN Annual Report 1976, p.57.

  18. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

  19. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  20. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

  1. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

  2. Training and research on the nuclear reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.

    1998-01-01

    The VR-1 training reactor is a light water reactor of the pool type using enriched uranium as the fuel. The moderator is demineralized light water, which also serves as the neutron reflector, biological shielding, and coolant. Heat evolved during the fission process is removed by natural convection. The reactor is used in the education of students in the field of reactor and neutron physics, dosimetry, nuclear safety, and instrumentation and control systems for nuclear facilities. Although primarily intended for students in various branches of technology (power engineering, nuclear engineering, physical engineering), this specialized facility is also used by students of faculties educating future natural scientists and teachers. Typical tasks trained at the VR-1 reactor include: measurement of delayed neutrons; examination of the effect of various materials on the reactivity of the reactor; measurement of the neutron flux density by various procedures; measurement of reactivity by various procedures; calibration of reactor control rods by various procedures; approaching the critical state; investigation of nuclear reactor dynamics; start-up, control and operation of a nuclear reactor; and investigation of the effect of a simulated nucleate boil on reactivity. In addition to the education of university-level students, training courses are also organized for specialists in the Czech nuclear programme

  3. Operational experience of the Marcoule reactors; Experience d'exploitation des reacteurs de Marcoule

    Energy Technology Data Exchange (ETDEWEB)

    Conte, F [Commissariat a l' Energie Atomique, Centre de Production de Plutonium, Marcoule (France). Centre d' Etudes Nucleaires

    1963-07-01

    The results obtaining from three years operation of the reactors G-2, G-3 have made it possible to accumulate a considerable amount of operational experience of these reactors. The main original points: - the pre-stressed concrete casing - the possibility of loading while under power - automatic temperature control have been perfectly justified by the results of operation. The author confirms the importance of these original solutions and draws conclusions concerning the study of future nuclear power stations. (author) [French] Les resultats atteints apres trois ans de fonctionnement des reacteurs G-2/G-3 permettent une accumulation considerable de l'experience d'exploitation de ces reacteurs. Les principales originalites: - caisson en beton precontraint - chargement en marche - surveillance automatique des temperatures sont largement justifiees par l'exploitation actuelle. L'auteur confirme l'interet de ces solutions d'avant-garde et en tire des conclusions pour les etudes de futures centrales nucleaires. (auteur)

  4. Reactor engineering department annual report. April 1, 1994 - March 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author).

  5. Reactor engineering department annual report. April 1, 1995 - March 31, 1996

    International Nuclear Information System (INIS)

    1996-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  6. Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

    International Nuclear Information System (INIS)

    1997-10-01

    This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department's programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

  7. Reactor engineering department annual report. April 1, 1994 - March 31, 1995

    International Nuclear Information System (INIS)

    1995-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  8. Reactor Engineering Department annual report. April 1, 1997 - March 31, 1998

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, Masaaki; Ohnuki, Akira; Ono, Toshihiko [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    1998-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1997 (April 1, 1997 - March 31, 1998). The major Department`s programs promoted in the year are the achievement of the world-strongest lasing of Free Electron Laser and the verification of the core thermal integrity during design basis events in PWRs. Other Major tasks of the Department are various basic researches on the advanced reactor system design studies, the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  9. Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-10-01

    This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department`s programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

  10. Reactor Engineering Department annual report. April 1, 1997 - March 31, 1998

    International Nuclear Information System (INIS)

    Ochiai, Masaaki; Ohnuki, Akira; Ono, Toshihiko

    1998-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1997 (April 1, 1997 - March 31, 1998). The major Department's programs promoted in the year are the achievement of the world-strongest lasing of Free Electron Laser and the verification of the core thermal integrity during design basis events in PWRs. Other Major tasks of the Department are various basic researches on the advanced reactor system design studies, the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  11. Reactor engineering department annual report. April 1, 1995 - March 31, 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  12. Irradiation experience with HTGR fuels in the Peach Bottom Reactor

    International Nuclear Information System (INIS)

    Scheffel, W.J.; Scott, C.B.

    1974-01-01

    Fuel performance in the Peach Bottom High-Temperature Gas-Cooled Reactor (HTGR) is reviewed, including (1) the driver elements in the second core and (2) the test elements designed to test fuel for larger HTGR plants. Core 2 of this reactor, which is operated by the Philadelphia Electric Company, performed reliably with an average nuclear steam supply availability of 85 percent since its startup in July 1970. Core 2 had accumulated a total of 897.5 equivalent full power days (EFPD), almost exactly its design life-time of 900 EFPD, when the plant was shut down permanently on October 31, 1974. Gaseous fission product release and the activity of the main circulating loop remained significantly below the limits allowed by the technical specifications and the levels observed during operation of Core 1. The low circulating activity and postirradiation examination of driver fuel elements have demonstrated the improved irradiation stability of the coated fuel particles in Core 2. Irradiation data obtained from these tests substantiate the performance predictions based on accelerated tests and complement the fuel design effort by providing irradiation data in the low neutron fluence region

  13. Experiences in controlling the upgrading of TRIGA 2000 Bandung reactor

    International Nuclear Information System (INIS)

    Huda, K.; Wibowo, Y.W.; Suprawhardana, M.S.

    2001-01-01

    TRIGA 2000 Bandung Reactor was established in 1961 for research, education and isotope production purposes. The reactor reached its first criticality in October 1964 and operated at nominal power of 250 kW until 1971. In 1971 the reactor was upgraded to the power level of 1000 kW. In order to raise the capacity of isotope production, the reactor has been upgraded again to the power level of 2000 kW. During the modification of the reactor, the Center for Research and Development of Nuclear Techniques (CRDNT) was management of the reactor as it faced many problems, either technical or non-technical ones. This caused the upgrading activities to take a long time. At this time, the reactor upgrading has almost finished, and the nuclear commissioning is going on. Several aspects and problems associated with the upgrading process have been reviewed and the results are discussed in the present paper. (author)

  14. IRPHE/B and W-SS-LATTICE, Spectral Shift Reactor Lattice Experiments

    International Nuclear Information System (INIS)

    2003-01-01

    Description: B and W has performed and analysed a series of physics experiments basically concerned with the technology of heterogeneous reactors moderated and cooled by a variable mixture of heavy and light water. A reactor so moderated is termed Spectral Shift Control Reactor (S SCR). In the practical application of this concept, the moderator mixture is rich in heavy water at the beginning of core life, so a relatively large fraction of the neutrons are epithermal and are absorbed in the fertile material. As fuel is consumed, the moderator is diluted with light water. In this way the neutron spectrum is shifted, thereby increasing the proportion of thermal neutrons and the reactivity of the system. The general objective of the S SCR Basic Physics Program was to study the nuclear properties of rod lattices moderated by D 2 O-H 2 O mixtures. The volume ratio of moderator to non-moderator in all lattices was approximately 1.0, and the fuel was either 4%-enriched UO 2 clad in stainless steel or 93%-enriched UO 2 -ThO 2 (Nth/N 15) pellets clad in aluminum. The D 2 O concentration in the moderator ranged from zero to about 90 mole %. The experimental program includes critical experiments with both types of fuel, exponential experiments at room temperature with both types of fuel, exponential experiments at elevated temperatures with the 4%-enriched UO 2 fuel, and neutron age measurements in ThO 2 lattices. The theoretical program included the development of calculation methods applicable to these systems, and the analysis and correlation of the experimental data. A first report provides the results of critical experiments performed under the Spectral Shift Control Reactor Basic Physics Program. A second report documents experimental results and theoretical interpretation of a series of twenty uniform lattice critical experiments in which the neutron spectrum is varied over a fairly broad range. A third report addresses issues that bear on the problems associated with

  15. Reactor Engineering Department annual report (April 1, 1986 - March 31, 1987)

    International Nuclear Information System (INIS)

    1987-08-01

    Research and development activities in the Department of Reactor Engineering in the fiscal year 1986 are described. The major activities of the Department are closely related to the reactor physics of very high temperature gas-cooled reactor, high conversion light water reactor and liquid metal fast breeder reactor and to blanket neutronics of fusion reactor. Contents of this report are divided into the activities on nuclear data and group constants, theoretical methods and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control, diagnosis and robotics. The activity of the Research Committee on Reactor Physics is also included. (author)

  16. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  17. Direct digital temperature control of the A-1 nuclear reactor

    International Nuclear Information System (INIS)

    Karpeta, C.

    1975-01-01

    The application is described of one of the modern control methods for designing an experimental digital temperature control system for heavy water moderated gas cooled reactors. The synthesis of the optimal stochastic regulator for reactor control in the area of the rated steady state was carried out using the method of dynamic programming and the Kalman filter technique. The analysis of the feedback circuit was conducted using control simulation on a universal digital computer. Results and experience are summed up. (author)

  18. International Reactor Physics Experiment Evaluation (IRPhE) Project. IRPhE Handbook - 2015 edition

    International Nuclear Information System (INIS)

    Bess, John D.; Gullifor, Jim

    2015-03-01

    The purpose of the International Reactor Physics Experiment Evaluation (IRPhE) Project is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. This work of the IRPhE Project is formally documented in the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments', a single source of verified and extensively peer-reviewed reactor physics benchmark measurements data. The evaluation process entails the following steps: Identify a comprehensive set of reactor physics experimental measurements data, Evaluate the data and quantify overall uncertainties through various types of sensitivity analysis to the extent possible, verify the data by reviewing original and subsequently revised documentation, and by talking with the experimenters or individuals who are familiar with the experimental facility, Compile the data into a standardized format, Perform calculations of each experiment with standard reactor physics codes where it would add information, Formally document the work into a single source of verified and peer reviewed reactor physics benchmark measurements data. The International Handbook of Evaluated Reactor Physics Benchmark Experiments contains reactor physics benchmark specifications that have been derived from experiments that were performed at nuclear facilities around the world. The benchmark specifications are intended for use by reactor designers, safety analysts and nuclear data evaluators to validate calculation techniques and data. Example calculations are presented; these do not constitute a validation or endorsement of the codes or cross-section data. The 2015 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments contains data from 143 experimental series that were

  19. Irradiation routine in the IPR-R1 Triga reactor

    International Nuclear Information System (INIS)

    Maretti Junior, F.

    1980-01-01

    Information about irradiations in the IPR-R1 TRIGA reactor and procedures necessary for radioisotope solicitation are presented All procedures necessary for asking irradiation in the reactor, shielding types, norms of terrestrial and aerial expeditions, payment conditions, and catalogue of disposable isotopes with their respective saturation activities are described. (M.C.K.)

  20. IEA-R1 reactor - Spent fuel management

    International Nuclear Information System (INIS)

    Mattos, J.R.L. De

    1996-01-01

    Brazil currently has one Swimming Pool Research Reactor (IEA-R1) at the Instituto de Pesquisas Energeticas e Nucleares - Sao Paulo. The spent fuel produced is stored both at the Reactor Pool Storage Compartment and at the Dry Well System. The present situation and future plans for spent fuel storage are described. (author). 3 refs, 2 figs, 2 tabs

  1. Thermohydraulic analysis for power increase of IEAR-1 reactor

    International Nuclear Information System (INIS)

    Umbehaun, Pedro E.; Bastos, Jose L.F.

    1996-01-01

    In this work has been presented the reactor core thermohydraulic model of IEAR-1, aiming its power operation increase from 2MW to 5MW. The design criteria adopted have been established in Safety Series 35. Three configurations of reactor core were analysed: fuel elements 20, 25 and 30

  2. Design of channel experiment equipment for measuring coolant velocity of innovative research reactor

    International Nuclear Information System (INIS)

    Muhammad Subekti; Endiah Puji Hastuti; Dedi Heriyanto

    2014-01-01

    The design of innovative high flux research reactor (RRI) requires high power so that the capability core cooling requires to be improved by designing the faster core coolant velocity near to the critical velocity limit. Hence, the critical coolant velocity as the one of the important parameter of the reactor safety shall be measured by special equipment to the velocity limit that may induce fuel element degradation. The research aims is to calculate theoretically the critical coolant velocity and to design the special experiment equipment namely EXNal for measuring the critical coolant velocity in fuel element subchannel of the RRI. EXNal design considers the critical velocity calculation result of 20.52 m/s to determine the variation of flow rate of 4.5-29.2 m 3 /h, in which the experiment could simulate the 1-4X standard coolant velocity of RSG-GAS as well as destructive test of RRI's fuel plate. (author)

  3. Experiment study on thermal mixing performance of HTR-PM reactor outlet

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Yangping, E-mail: zhouyp@mail.tsinghua.edu.cn [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China); Hao, Pengfei [School of Aerospace, Tsinghua University, Beijing 100084 (China); Li, Fu; Shi, Lei [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China); He, Feng [School of Aerospace, Tsinghua University, Beijing 100084 (China); Dong, Yujie; Zhang, Zuoyi [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China)

    2016-09-15

    A model experiment is proposed to investigate the thermal mixing performance of HTR-PM reactor outlet. The design of the test facility is introduced, which is set at a scale of 1:2.5 comparing with the design of thermal mixing structure at HTR-PM reactor outlet. The test facility using air as its flow media includes inlet pipe system, electric heaters, main mixing structure, hot gas duct, exhaust pipe system and I&C system. Experiments are conducted on the test facility and the values of thermal-fluid parameters are collected and analyzed, which include the temperature, pressure and velocity of the flow as well as the temperature of the tube wall. The analysis results show the mixing efficiency of the test facility is higher than that required by the steam generator of HTR-PM, which indicates that the thermal mixing structure of HTR-PM fulfills its design requirement.

  4. Plutonium Recycle Test Reactor (PRTR). Operating Experience and Supporting R and D, Its Application to Heavy-Water Power Reactor Design and Operation

    Energy Technology Data Exchange (ETDEWEB)

    Harty, H. [Battelle Memorial Institute, Pacific Northwest Laboratories, Richland, WA (United States)

    1968-04-15

    Convincing answers to questions about heavy-water, pressure-tube, power reactors, e.g. pressure-tube serviceability, heavy-water management problems, long-term behaviour of special pressure-tube reactor components, and unique operating maintenance problems (compared to light-water reactors) must be based on actual operating experience with that type of reactor. PRTR operating experience and supporting R and D studies, although not always simple extrapolations to power reactors, can be summarized in a context applicable to future heavy-water power reactors, as follows: 1. Pressure-tube life, in a practical case, need not be limited by creep, gross hydriding, corrosion, or mechanical damage. The possibility that growth of a defect (perhaps service-induced) to a size that is critical under certain operating conditions, remains a primary unknown in pressure- tube life extrapolations. A pressure-tube failure in PRTR (combined with gross release of fuel material) proved only slightly more inconvenient, time consuming, and damaging to the reactor proper, than occurred with a gross failure of a fuel element in PRTR. 2. Routine operating losses of heavy water appear tractable in heavy-water-cooled power reactors; losses from low-pressure systems can be insignificant over the life of a plant. Non-routine losses may prove to be the largest component of loss over the life of a plant. 3. The performance of special components in PRTR, e.g. the calandria and shields, has not deteriorated despite being subjected to non-standard operating conditions. The calandria now contains a light-water reflector with single barrier separation from the heavy-water moderator. The carbon steel shields (containing carbon steel shot) show no deterioration based on pressure drop measurements and piping activation immediately outside the shields. The helium pressurization system (for primary coolant pressurization) remains a high maintenance system, and cannot be recommended for power reactors, based

  5. (CSTR) and 1-plug flow reactor (PFR)

    African Journals Online (AJOL)

    Administrator

    2010-12-27

    Dec 27, 2010 ... 2Department of Chemical Engineering, Obafemi Awolowo University, Ile Ife, Osun State, Nigeria. ... modeled as PFR and described by Michaelis. Menten equation. Designed equations derived from the two equations are used for the reactor sizing of ... Herbivores make a living on cellulose by possessing.

  6. Use of self powered neutron detectors in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Galo Rocha, F. del.

    1989-01-01

    A survey of self-powered neutron detectors, SPND, which are used as part of the in-core instrumentation of nuclear reactors is presented. Measurements with Co and Er SPND's were made in the IEA-R1 reactor for determining the neutron flux distribution and the integral reactor power. Due to the size of the available detectors, the neutron flux distribution could not be obtained with accuracy. The results obtained in the reactor power measurements demonstrate that the SPND have the linearity and the quick response necessary for a reactor power channel. This work also presents a proposed design of a SPND using Pt as wire emissor. This proposed design is based in the experience gained in building two prototypes. The greatest difficulties encountered include materials and technology to perform the delicate weldings. (author)

  7. Criticality calculations in reactor accelerator coupling experiment (Race)

    International Nuclear Information System (INIS)

    Reda, M.A.; Spaulding, R.; Hunt, A.; Harmon, J.F.; Beller, D.E.

    2005-01-01

    A Reactor Accelerator Coupling Experiment (RACE) is to be performed at the Idaho State University Idaho Accelerator Center (IAC). The electron accelerator is used to generate neutrons by inducing Bremsstrahlung photon-neutron reactions in a Tungsten- Copper target. This accelerator/target system produces a source of ∼1012 n/s, which can initiate fission reactions in the subcritical system. This coupling experiment between a 40-MeV electron accelerator and a subcritical system will allow us to predict and measure coupling efficiency, reactivity, and multiplication. In this paper, the results of the criticality and multiplication calculations, which were carried out using the Monte Carlo radiation transport code MCNPX, for different coupling design options are presented. The fuel plate arrangements and the surrounding tank dimensions have been optimized. Criticality using graphite instead of water for reflector/moderator outside of the core region has been studied. The RACE configuration at the IAC will have a criticality (k-effective) of about 0,92 and a multiplication of about 10. (authors)

  8. Liquid nitrogen - water interaction experiments for fusion reactor accident scenarios

    International Nuclear Information System (INIS)

    Duckworth, R.; Murphy, J.; Pfotenhauer, J.; Corradini, M.

    2001-01-01

    With the implementation of superconducting magnets in fusion reactors, the possibility exists for the interaction between water and cryogenic systems. The interaction between liquid nitrogen and water was investigated experimentally and numerically. The rate of pressurization and peak pressure were found to be driven thermodynamically by the expansion of the water and the boil-off of the liquid nitrogen and did not have a vapor explosion nature. Since the peak pressure was small in comparison to previous work with stratified geometries, the role of the geometry of the interacting fluids has been shown to be significant. Comparisons of the peak pressure and the rate of pressurization with respect to the ratio of the liquid nitrogen mass to water mass reveal no functional dependence as was observed in the liquid helium-water experiments. A simple thermodynamic model provides a fairly good description of the pressure rise data. From the data, the model will allow one to extract the interaction area of the water. As with previous liquid helium-water interaction experiments, more extensive investigation of the mass ratio and interaction geometry is needed to define boundaries between explosive and non-explosive conditions. (authors)

  9. Natural-draught cooling tower of the Philippsburg-1 reactor

    International Nuclear Information System (INIS)

    Ernst, G.; Wurz, D.

    1983-01-01

    In spring 1980 a comprehensive research programm was carried out on the natural-draught cooling tower of the Philippsburg-1 reactor. The study was meant to synchronously acquire all parameters necessary for the evaluation of plant operation and cooling tower emissions. The study is subdivided into 8 sub-projects. Parts 1 to 7 that are included in this progress-of-work report describe experimental work and discuss the results. A critical analysis of measuring results proves that the values for operational behaviour and cooling tower emissions were duly anticipated. Even a very critical judgment of the results can exclude direct or indirect hazards for humans, animals and plants owing to cooling tower emissions. Sub-project 8 compares results from diffusion calculations (24 models) to results gained from experiments. The results of sub-project 8 will be published in a progress report to come. (orig.) [de

  10. The KASKA project - a Japanese medium-baseline reactor-neutrino oscillation experiment to measure the mixing angle $\\theta_{13}$ -

    OpenAIRE

    Kuze, Masahiro; Collaboration, for the KASKA

    2005-01-01

    A new reactor-neutrino oscillation experiment, KASKA, is proposed to measure the unknown neutrino-mixing angle $\\theta_{13}$ using the world's most powerful Kashiwazaki-Kariwa nuclear power station. It will measure a very small deficit of reactor-neutrino flux using three identical detectors, two placed just close to the sources and one at a distance of about 1.8km. Its conceptual design and physics reach are discussed.

  11. French experience in design, operation and revamping of nuclear research reactors, in support of advanced reactors development

    International Nuclear Information System (INIS)

    Barre, B.; Bergeonneau, P.; Merchie, F.; Minguet, J.L.; Rousselle, P.

    1996-01-01

    The French nuclear program is strongly based on the R and D work performed in the CEA nuclear research centers and particularly on the various experimental programs carried out in its research reactors in the frame of cooperative actions between the Commissariat a l'Energie Atomique (CEA), Framatome and Electricite de France (EDF). Several types of research reactors have been built by Technicatome and CEA to carry out successfully this considerable R and D work on fuels and materials, among them the socalled Materials Testing Reactors (MTR) SILOE (35 MW) and OSIRIS (70 MW) which are indeed very well suited for technological irradiations. Their simple and flexible design and the large irradiation space available around the core, the SILOE and OSIRIS reactors can be shared by several types of applications such as fuel and material testings for nuclear power plants, radioisotopes production, silicon doping and fundamental research. It is worthwhile recalling that Technicatome and CEA have also built research reactors fully dedicated to safety experimental studies, such as the CABRI, SCARABEE and PHEBUS reactors at Cadarache, and others dedicated to fundamental research such as ORPHEE (14 MW) and the Reacteur a Haut Flux -High Flux Reactor- (RHF 57 MW). This paper will present some of the most significant conceptual and design features of all these reactors as well as the main improvements brought to most of them in the last years. Based on this wide experience, CEA and Technicatome have specially designed for export a new multipurpose research reactor named SIRIUS, with two versions depending on the utilization spectrum and the power range (5 MW to 30 MW). At last, CEA has recently launched the preliminary project study of a new MTR, the Jules Horowitz Reactor, to meet the future needs of fuels and materials irradiations in the next 4 or 5 decades, in support of the French long term nuclear power program. (J.P.N.)

  12. RA research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1985

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1985-01-01

    According to the plan, RA reactor was to be in operation in mid September 1985. But, since the building of the emergency cooling system, nor the reconstruction of the existing special ventilation system were not finished until the end of August reactor was not operated during 1985. During the previous four years reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care, which was cancelled in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981-1984. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks have started: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. IAEA has approved the amount of 1,300,000 US dollars for the renewal of the instrumentation [sr

  13. Data sheets of fission product release experiments for light water reactor fuel, (2)

    International Nuclear Information System (INIS)

    Ishiwatari, Nasumi; Nagai, Hitoshi; Takeda, Tsuneo; Yamamoto, Katsumune; Nakazaki, Chozaburo.

    1979-07-01

    This is the second data sheets of fission products (FP) release experiments for light water reactor fuel. Results of five FP release experiments from the third to the seventh are presented: results of pre-examinations of UO 2 pellets, photographs of parts of fuel rod assemblies for irradiation and the assemblies, operational conditions of JMTR and OWL-1, variations of radioiodine-131 level in the main loop coolant during experimental periods, and representative results of post-irradiation examinations of respective fuel rods. (author)

  14. Safety Research Experiment Facility Project. Conceptual design report. Volume VII. Reactor cooling

    International Nuclear Information System (INIS)

    1975-12-01

    The Reactor Cooling System (RCS) will provide the required cooling during test operations of the Safety Research Experiment Facility (SAREF) reactor. The RCS transfers the reactor energy generated in the core to a closed-loop water storage system located completely inside the reactor containment building. After the reactor core has cooled to a safe level, the stored heat is rejected through intermediate heat exchangers to a common forced-draft evaporative cooling tower. The RCS is comprised of three independent cooling loops of which any two can remove sufficient heat from the core to prevent structural damage to the system components

  15. FELIX experiments and computational needs for eddy current analysis of fusion reactors

    International Nuclear Information System (INIS)

    Turner, L.R.

    1984-01-01

    In a fusion reactor, changing magnetic fields are closely coupled to the electrically-conducting metal structure. This coupling is particularly pronounced in a tokamak reactor in which magnetic fields are used to confine, stabilize, drive, and heat the plasma. Electromagnetic effects in future fusion reactors will have far-reaching implications in the configuration, operation, and maintenance of the reactors. This paper describes the impact of eddy-current effects on future reactors, the requirements of computer codes for analyzing those effects, and the FELIX experiments which will provide needed data for code validation

  16. Determination of the neutron flux for the Yankee Rowe experiment in the Ford Nuclear Reactor

    International Nuclear Information System (INIS)

    Cacciapouti, R.J.; Petrusha, L.

    1994-01-01

    Yankee Atomic Electric Company undertook a Test Irradiation Program at the Ford Nuclear Reactor of the University of Michigan. The program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials. The program was also intended to remove uncertainties in the existing reactor vessel fluence and damage predictions on the Yankee Rowe reactor vessel steel. Since this is the first in-core experiment of this type for the Ford Nuclear Reactor, the measurement of the reaction rate and the estimate of the fluence are presented

  17. Study on the Neutrino Oscillation with a Next Generation Medium-Baseline Reactor Experiment

    International Nuclear Information System (INIS)

    Joo, Kyung Kwang; Shin, Chang Dong

    2014-01-01

    For over fifty years, reactor experiments have played an important role in neutrino physics, in both discoveries and precision measurements. One of the methods to verify the existence of neutrino is the observation of neutrino oscillation phenomena. Electron antineutrinos emitted from a reactor provide the measurement of the small mixing angle θ 13 , providing rich programs of neutrino properties, detector development, nuclear monitoring, and application. Using reactor neutrinos, future reactor neutrino experiments, more precise measurements of θ 12 ,Δm 12 2 , and mass hierarchy will be explored. The precise measurement of θ 13 would be crucial for measuring the CP violation parameters at accelerators. Therefore, reactor neutrino physics will assist in the complete understanding of the fundamental nature and implications of neutrino masses and mixing. In this paper, we investigated several characteristics of RENO-50, which is a future medium-baseline reactor neutrino oscillation experiment, by using the GloBES simulation package

  18. Reactor physics experiments related to transmutation in the KUCA

    Energy Technology Data Exchange (ETDEWEB)

    Shiroya, Seiji [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst.

    1997-11-01

    At the Kyoto University Critical Assembly (KUCA), {sup 237}Np/{sup 235}U fission rate ratios are being measured using the back-to-back type double fission chamber to examine the nuclear data and the computational method for the transmutation of minor actinides (MA) in light water reactors (LWRs). The neutron spectra of cores are systematically being varied by changing the moderator-to-fuel volume ratio (V{sub m}/V{sub f}). The measured data are being compared with the calculated results by SRAC with three different nuclear data files. It has been indicated that the calculated results with JENDL-3.2 agreed better with the measured ones than those with JENDL-3.1 and ENDF/B-VI, although the calculated results underestimated the measured ones by around 10%. (author)

  19. FAFTRCS: an experiment in computerized reactor safety systems

    International Nuclear Information System (INIS)

    Chisholm, G.H.

    1985-01-01

    Nuclear Power Plant availability and reliability could be improved by the integration of computers into the control environment. However, computer-based systems are historically viewed as being unreliable. This places a burden upon the designer to demonstrate adequate reliability and availability for the computer. The complexity associated with computers coupled with the manual nature of these demonstrations results in a high cost which typically has been justified for critical applications only. This paper investigates a methodology for automating this process and discusses a project which intends to apply this methodology to design verification and validation for a control system which will be installed and tested in an actual reactor control environment. 7 refs., 4 figs., 1 tab

  20. Nuclear engineering laboratory self regulated power oscillation experiments at the Health Physics Research Reactor

    International Nuclear Information System (INIS)

    Miller, L.F.; Mihalczo, J.T.; Bailiff, E.G.; Woody, N.D.; Gardner, G.D.

    1983-01-01

    Self regulated power oscillation experiments with a variety of initial conditions have been performed with the ORNL Health Physics Research Reactor (HPRR) by undergraduate nuclear engineering students from The University of Tennessee for several years. These experiments demonstrate the coupling between reactor kinetics and heat transfer and show how the temperature coefficient of reactivity affects reactor behavior. A model that consists of several coupled first order nonlinear differential equations is used to calculate the temperature of the core center and surface and power as a function of time which are compared with the experimental data; also, the model is also used to study the effects of various model parameters and initial conditions on the amplitude, frequency and damping of the power and temperature oscillations. A previous paper presented some limited experimental results and demonstrated the correspondence between a simple point model and the experimental data. This paper presents the results of experiments for: (1) the initial power fixed at 9 kW with central core temperatures of 300 0 F and 500 0 F, annd (2) the initial central core temperature fixed at 500 0 F with initial powers of 6 and 8 kW

  1. The functioning of the reactors G2-G3 at Marcoule and E.D.F. 1; Experience de fonctionnement des reacteurs G2-G3 de Marcoule et enseignements des essais de demarrage du reacteur E.D.F. 1 de Chinon

    Energy Technology Data Exchange (ETDEWEB)

    Boussard, R; Conte, F [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires; Stolz, J M [Electricite de France (EDF), 75 - Paris (France)

    1964-07-01

    After resuming briefly the characteristics of the installations G2-G3 at Marcoule and EDF 1 at Chinon, the authors review the main aspects of the tests, the starting and the exploitation of these reactors. Among the various points examined, particular emphasis is given to the devices of original nature such as tubular fuel elements, flattening of the neutron flux by stuffing, behaviour of the reactor tanks and the cooling circuits, the blowers, unloading devices, regulation and functioning of the informations. This analysis deals equally with the performances obtained and the difficulties and the various incidents experienced during the initial starting period. Among the more interesting results, the progressive increase in the power of the Marcoule reactors is mentioned, obtained through a better knowledge of the parameters covering the functioning of the reactors such as the distribution of the flux and the temperatures etc... acquired during the course of the exploitation of the reactor. The conclusion reached by the authors is that the experience gained on these installations has shown: - that during an initial period, adjustments became necessary, all of which turned out to be possible, - that an analysis of their functioning has permitted the progressive movement towards a truly industrial exploitation. (authors) [French] Les auteurs, apres un bref rappel des caracteristiques des installations G2 - G3 de MARCOULE et E.D.F. 1 de CHINON, passent en revue les principaux aspects des essais, de la mise en service et de l'exploitation de ces centrales. Parmi les divers points examines, une attention speciale est accordee aux dispositifs presentant un caractere original tels que elements combustibles tubulaires, aplatissement du flux neutronique par gavage, comportement des caissons des reacteurs et des circuits de refroidissement, soufflantes, appareils de dechargement, regulation et fonctionnement des informations. L'analyse presentee porte tant sur les

  2. Preliminary Study of 20 MWth Experiment Power Reactor based on Pebble Bed Reactor

    Science.gov (United States)

    Irwanto, Dwi; Permana, Sidik; Pramuditya, Syeilendra

    2017-07-01

    In this study, preliminary design calculations for experimental small power reactor (20 MWt) based on Pebble Bed Reactor (PBR) are performed. PBR technology chosen due to its advantages in neutronic and safety aspects. Several important parameters, such as fissile enrichment, number of fuel passes, burnup and effective multiplication factor are taken into account in the calculation to find neutronic characteristics of the present reactor design.

  3. Fuel Element Experience at the Halden Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aas, S. [OECD Halden Reactor Project, Halden (Norway); Videm, K.; Hanevik, A. [Institutt for Atomenergi, Kjeller (Norway)

    1968-04-15

    The penalty for neutron absorbing materials is higher for a reactor moderated with heavy water than one with light water. As Zircaloy and enriched uranium were not readily available in 1954 when the design of the first fuel charge for HBWR was frozen, fuel elements of natural uranium metal clad in a specially developed aluminium alloy (A 1 0.3% Fe, 0.03% Si) were used. The temperature was limited to 150 Degree-Sign C and with this limitation the general behaviour of the elements was good. In I960, in another effort to maintain a good neutron economy, a couple of elements with as thin cladding as 0.25 mm A1S1 316, stainless steel with an unsegmented length of 2 m supported by wire grid spacers were tested. These elements with 1.5% enriched UO{sub 2} behaved satisfactorily at 150'C. Elements of a rather similar construction failed due to stress corrosion during the later operation at 230 'C. The reason for the different behaviour is probably the higher stresses in the cladding, due to the increased pressure, possibly combined with a short period with a high chloride content in the heavy water. The second fuel core with 1.5% enriched UO{sub 2} clad in Zircaloy-2 was installed in order to permit an increase in temperature to 230 Degree-Sign C and in power from 5 to 20 MW(th). The maximum burnup obtained is 11000 MWd/t and the maximum heat rating 375 W/cm with no fracture failure and practically no change in appearance according to the post-irradiation examination. One element was deliberately taken to burn-out conditions by throttling the water flow. After a series of burn-outs, the element finally failed because of over-temperature. The successful use of aluminium cladding at 150 Degree-Sign C mitiated an effort for making aluminium alloys suitable for normal power reactor operation. Promising properties were found for an alloy (designated IFA 3 aluminium) with A1 10% Si, 1% Ni, 1% Mg, 0.3% Fe + Ti. Despite increase in corrosion rate under heat transfer conditions

  4. GROWTH OF THE INTERNATIONAL CRITICALITY SAFETY AND REACTOR PHYSICS EXPERIMENT EVALUATION PROJECTS

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; John D. Bess; Jim Gulliford

    2011-09-01

    Since the International Conference on Nuclear Criticality Safety (ICNC) 2007, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have continued to expand their efforts and broaden their scope. Eighteen countries participated on the ICSBEP in 2007. Now, there are 20, with recent contributions from Sweden and Argentina. The IRPhEP has also expanded from eight contributing countries in 2007 to 16 in 2011. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments1' have increased from 442 evaluations (38000 pages), containing benchmark specifications for 3955 critical or subcritical configurations to 516 evaluations (nearly 55000 pages), containing benchmark specifications for 4405 critical or subcritical configurations in the 2010 Edition of the ICSBEP Handbook. The contents of the Handbook have also increased from 21 to 24 criticality-alarm-placement/shielding configurations with multiple dose points for each, and from 20 to 200 configurations categorized as fundamental physics measurements relevant to criticality safety applications. Approximately 25 new evaluations and 150 additional configurations are expected to be added to the 2011 edition of the Handbook. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments2' have increased from 16 different experimental series that were performed at 12 different reactor facilities to 53 experimental series that were performed at 30 different reactor facilities in the 2011 edition of the Handbook. Considerable effort has also been made to improve the functionality of the searchable database, DICE (Database for the International Criticality Benchmark Evaluation Project) and verify the accuracy of the data contained therein. DICE will be discussed in separate papers at ICNC 2011. The status of the

  5. A review of experiments and results from the transient reactor test (TREAT) facility

    International Nuclear Information System (INIS)

    Deitrich, L. W.

    1998-01-01

    The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop

  6. Neutron density optimal control of A-1 reactor analoque model

    International Nuclear Information System (INIS)

    Grof, V.

    1975-01-01

    Two applications are described of the optimal control of a reactor analog model. Both cases consider the control of neutron density. Control loops containing the on-line controlled process, the reactor of the first Czechoslovak nuclear power plant A-1, are simulated on an analog computer. Two versions of the optimal control algorithm are derived using modern control theory (Pontryagin's maximum principle, the calculus of variations, and Kalman's estimation theory), the minimum time performance index, and the quadratic performance index. The results of the optimal control analysis are compared with the A-1 reactor conventional control. (author)

  7. Experience in handling core subassemblies in sodium cooled reactor KNK and test rigs

    International Nuclear Information System (INIS)

    Althaus; Jansing; Kesseler; Kirchner; Menck

    1974-01-01

    Compared with a water cooled reactor plant a sodium cooled reactor plant presents a number of problems which result from the specific nature of sodium. These problems that must be faced during all handling operations are mainly: 1. The rapid reaction of sodium in air requires handling to be done only under cover gas. 2. The temperature of all sodium-wetted components is to be kept above the melting point of sodium. 3. Poor draining of removed reactor components due to the high surface tension of sodium and the associated danger of dripping radioactive sodium may produce radiation or contamination problems. 4. Sodium is not transparent. The sum of these and further influences dictate that the general handling usually is carried out without visual means, though a method is under development in the USA to use ultrasonic for under sodium 'viewing'. These limitations to sodium component handling are applicable to all sodium reactor plants, several of which are discussed in this report. After the description of the handling systems of the KNK plant now operating at Karlsruhe, the experience with the SNR test rig and finally the handling systems for SNR 300 and SNR 2 are discussed

  8. Sodium coolant of fast reactors: Experience and problems

    International Nuclear Information System (INIS)

    Kozlov, F.A.; Volchkov, L.G.; Drobyshev, A.V.; Nikulin, M.P.; Kochetkov, L.A.; Alexeev, V.V.

    1997-01-01

    In present report the following subjects are considered: state of the coolant and sodium systems under normal operating condition as well as under decommissioning, disclosing of sodium circuits and liquidation of its consequences, cleaning from sodium and decontamination under repairing works of equipment and circuits. Cleaning of coolant and sodium systems under normal operating conditions and under accident contamination. Cleaning of the equipment under repairing works and during decommissioning from sodium and products of its interaction with water and air. Treatment of sodium waste, taking into account a possibility of sodium fires. It is shown that the state of coolant, cover gas, surfaces of constructive materials which are in contact with them, cleaning systems, formed during installation operation require development of specific technologies. Developed technologies ensured safety operation of sodium cooled installations as in normal operating conditions so in abnormal situations. R and D activities in this field and experience gained provided a solid base for coping with problems arising during decommissioning. Prospective research problems are emphasized where the future efforts should be concentrated in order to improve characteristics of sodium cooled reactors and to make their decommissioning optimal and safe. (author)

  9. Technical report on operating experience with boiling water reactor offgas systems

    International Nuclear Information System (INIS)

    Lo, R.; Barrett, L.; Grimes, B.; Eisenhut, D.

    1978-03-01

    Over 100 reactor years of Boiling Water Reactor (BWR) operating experience have been accumulated since the first commercial operation of BWRs. A number of incidents have occurred involving the ''offgas'' of these Boiling Water Reactors. This report describes the generation and processing of ''offgas'' in Boiling Water Reactors, the safety considerations regarding systems processing the ''offgas'', operating experience involving ignitions or explosions of ''offgas'' and possible measures to reduce the likelihood of future ignitions or explosions and to mitigate the consequences of such incidents should they occur

  10. Simulator experiments: effects of experience of senior reactor operators and of presence of a shift technical advisor on performance in a boiling water reactor control room

    International Nuclear Information System (INIS)

    Beare, A.N.; Dorris, R.E.; Gray, L.H.

    1984-12-01

    This report describes the first experiment in a Nuclear Regulatory Commission-sponsored program of training simulator experiments and field data collection to evaluate the effects of selected performance shaping factors on the performance of nuclear power plant control room operators. The factors investigated were the experience level of the Senior Reactor Operator (SRO) and the presence of a Shift Technical Advisor (STA). Data were collected from 16 two-man crews of licensed operators (one SRO and one RO). The crews were split into high and low SRO-experience groups on the basis of the years of experience of the SROs as SROs. One half (4 of the 8 crews in each group) of the high- and low-SRO experience groups were assisted by an STA or an SRO acting as an STA. The crews responded to four simulated plant casualties which ranged in severity from an uncomplicated turbine trip to an anticipated transient without scram (ATWS). No significant differences in overall performance were found between groups led by high (25 to 114 months licensed as an SRO) and low (1 to 17 months as an SRO) experience SROs. However, crews led by low experience SROs tended to have shorter task performance times than crews led by high experience SROs. Although a tendency for the STA-assisted groups to score higher on four of the five measures was observed, the presence of the STA had no statistically significant effect on overall team performance. The correlation between individual performance, as measured by four of the task performance measures, and experience, measured by months as a licensed operator, was not statistically significant, nor was the correlation between task performance and recency of simulator training. 18 references, 5 figures, 13 tables

  11. Licensing of the first reload of Angra-1 reactor

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.

    1985-01-01

    The historical aspects related to the licensing of the first reload of Angra-1 reactor are presented. The dates, the institutions, the experts, as well as the documents generated during that process are presented. (M.I.)

  12. RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, Stephanie; Van-Duysen, Jean Claude

    2005-01-01

    Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called 'Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, ...) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program

  13. Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

    Directory of Open Access Journals (Sweden)

    Krása Antonín

    2017-01-01

    Full Text Available VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector. Discrepancies between experiments and Monte Carlo calculations (MCNP5 of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2 are presented.

  14. Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

    Science.gov (United States)

    Krása, Antonín; Kochetkov, Anatoly; Baeten, Peter; Vittiglio, Guido; Wagemans, Jan; Bécares, Vicente

    2017-09-01

    VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector). Discrepancies between experiments and Monte Carlo calculations (MCNP5) of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler) depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2) are presented.

  15. Irradiation creep experiments on fusion reactor candidate structural materials

    International Nuclear Information System (INIS)

    Hausen, H.; Cundy, M.R.; Schuele, W.

    1991-01-01

    Irradiation creep rates were determined for annealed and cold-worked AMCR- and 316-type steel alloys in the high flux reactor at Petten, for various irradiation temperatures, stresses and for neutron doses up to 4 dpa. Primary creep elongations were found in all annealed materials. A negative creep elongation was found in cold-worked materials for stresses equal to or below about 100 MPa. An increase of the negative creep elongation is found for decreasing irradiation temperatures and decreasing applied stresses. The stress exponent of the irradiation creep rate in annealed and cold-worked AMCR alloys is n = 1.85 and n = 1.1, respectively. The creep rates of cold-worked AMCR alloys are almost temperature independent over the range investigated (573-693 K). The results obtained in the HFR at Petten are compared with those obtained in ORR and EBR II. The smallest creep rates are found for cold-worked materials of AMCR- and US-PCA-type at Petten which are about a factor two smaller than the creep rates obtained of US-316 at Petten or for US-PCA at ORR or for 316L at EBR II. The scatter band factor for US-PCA, 316L, US-316 irradiated in ORR and EBR II is about 1.5 after a temperature and damage rate normalization

  16. Control-rod interference effects observed during reactor physics experiments with nuclear ship 'MUTSU'

    International Nuclear Information System (INIS)

    Itagaki, Masafumi; Miyoshi, Yoshinori; Gakuhari, Kazuhiko; Okada, Noboru; Sakai, Tomohiro.

    1993-01-01

    The control rods in the reactor of the nuclear ship MUTSU are classified into four groups: groups G1 and G2 are located in the central part of the core, while groups G3 and G4 are in the peripheral zone of the core. Several types of mutual interference effects among these control-rod groups were observed during reactor physics experiments with this reactor. During normal hot operations, positive shadowing was dominant between the G1 and G2 groups; the degree of the shadowing effect of one rod group depended on the position of the other rod group. Both positive and negative shadowing effects occurred between an inner rod group (G1 or G2) and an outer group (G3 or G4) depending on the three-dimensional arrangement of the control rods. The rod worths of G1 and G2 increased as a result of slight core burnup, about 1,400 MWd/t, mainly due to the decrease in shadowing effects resulting from a change in control-rod pattern. A three-dimensional diffusion calculation with internal control-rod boundary conditions has proved to be useful for analyzing these various interaction effects. (author)

  17. Reactor Section standard analytical methods. Part 1

    Energy Technology Data Exchange (ETDEWEB)

    Sowden, D.

    1954-07-01

    the Standard Analytical Methods manual was prepared for the purpose of consolidating and standardizing all current analytical methods and procedures used in the Reactor Section for routine chemical analyses. All procedures are established in accordance with accepted practice and the general analytical methods specified by the Engineering Department. These procedures are specifically adapted to the requirements of the water treatment process and related operations. The methods included in this manual are organized alphabetically within the following five sections which correspond to the various phases of the analytical control program in which these analyses are to be used: water analyses, essential material analyses, cotton plug analyses boiler water analyses, and miscellaneous control analyses.

  18. Fast reactor operating experience gained in Russia: Analysis of anomalies and abnormal operation cases

    International Nuclear Information System (INIS)

    Ashurko, Y.M.; Baklushin, R.P.; Zagorulko, Y.I.; Ivanenko, V.N.; Matveyev, V.P.; Vasilyev, B.A.

    2000-01-01

    Review of various anomalous events and abnormal operation experience gained in the process of Russian fast reactors operation is given in the paper. The main information refers to the BN-600 demonstration reactor operation. Statistical data on sodium leaks and steam generator failures are presented, and sources of these events and countermeasures taken to avoid their appearance on the operating reactors as well as related changes made in the BN-800 reactor design are considered. In the paper, some features of impurities behaviour are considered in various modes of the BN-600 reactor operation. Information is given on the impurities ingress into the circuits, on abnormal situation emerged in the process of the BN-600 reactor operation and its probable cause. Information is presented on the event related to the increased torque of the BN-600 reactor central rotating column and repair works performed. (author)

  19. Development, operational experience and implications for future design of fast reactors in Western Europe

    International Nuclear Information System (INIS)

    Brandstetter, A.; Broomfield, A.M.; Saitcevsky, B.

    1990-01-01

    Over the past 30 years, the partners now collaborating in Europe on fast-reactor development have taken the technology from a theoretical possibility to an engineering reality. In that time there has been a progression from experimental zero-energy facilities followed by small power-producing engineering test reactors, to prototype reactors and a large commercial-size reactor. The paper describes the highlights of the reactor programmes in the partner countries and by example illustrates the experience gained from reactor operation and some of the principal activities in the supporting development programme. These include such topics as fuel performance, fast-neutron physics, liquid-metal thermal hydraulics, sodium chemistry, instrumentation and safety aspects. The paper concludes by summarizing some of the main objectives of the current development programme, which is directed to the support of the European Fast Reactor design being prepared by the European design and construction companies. (author)

  20. New measuring and protection system at VR-1 training reactor

    International Nuclear Information System (INIS)

    Kropik, M.; Jurickova, M.

    2006-01-01

    The contribution describes the new measuring and protection system of the VR-1 training reactor. The measuring and protection system upgrade is an integral part of the reactor I and C upgrade. The new measuring and protection system of the VR-1 reactor consists of the operational power measuring and the independent power protection systems. Both systems measure the reactor power and power rate, initiate safety action if safety limits are exceeded and send data (power, power rate, status, etc.) to the reactor control system. The operational power measuring system is a full power range system that receives signal from a fission chamber. The signal is evaluated according to the reactor power either in the pulse or current mode. The current mode utilizes the DC current and Campbell techniques. The new independent power protection system operates in the two highest reactor power decades. It receives signals from a boron chamber and evaluates it in the pulse mode. Both systems are computer based. The operational power measuring and independent power protection systems are diverse - different types and location of chambers, completely different hardware, software algorithms for the power and power rate calculations, software development tools and teems for the software manufacturing. (author)

  1. Modernization and Refurbishment of the RECH-1 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Daie, J. [Nuclear Application Department, Chilean Nuclear Energy Commission (CCHEN), Santiago (Chile)

    2014-08-15

    The Chilean Nuclear Energy Commission (Comisión Chilena de Energía Nuclear, or CCHEN) has operated the RECH-1 research reactor since 1974. This reactor is located at La Reina Nuclear Centre in Santiago, Chile. It is a pool type reactor using LEU MTR fuel assemblies, light water as moderator and coolant, and beryllium as reflector. The reactor has been operated at the nominal power of 5 MW in a continuous shift of 20 hours per week, 48 weeks per year. The main utilizations of the RECH-1 reactor are radioisotope production and neutron activation analysis. Among the most relevant refurbishment and modernization campaigns undertaken at the reactor are: full core conversion to the use of LEU fuel, replacement of the cooling tower, improvement of the containment building by changing the doors and gates and by a better sealant for the penetrations, introduction of an additional source of water by connecting the raw water supply system to the emergency cooling system, improvement of the emergency ventilation system, introduction of a fire detection and alarm system for detection and mitigation to protect the I&C racks, introduction of a radioactive liquid release for those generated at the reactor, introduction of a delay tank degasification system and renewal of the environmental monitoring system. At present we are assessing the possibility of replacing the old analog electronics of control for new digital systems. Detailed descriptions of these diverse activities are presented in the paper. (author)

  2. Compact tokamak reactors. Part 1 (analytic results)

    International Nuclear Information System (INIS)

    Wootton, A.J.; Wiley, J.C.; Edmonds, P.H.; Ross, D.W.

    1996-01-01

    We discuss the possible use of tokamaks for thermonuclear power plants, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First we review and summarize the existing literature. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamak power plant, by including the power required to drive the toroidal field, and considering two extremes of plasma current drive efficiency. The analytic results will be augmented by a numerical calculation which permits arbitrary plasma current drive efficiency; the results of which will be presented in Part II. Third, a scaling from any given reference reactor design to a copper toroidal field coil device is discussed. Throughout the paper the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculating electric power. We conclude that the latest published reactor studies, which show little advantage in using low aspect ratio unless remarkably high efficiency plasma current drive and low safety factor are combined, can be reproduced with the analytic model

  3. Study of dietary supplements compositions by neutron activation analysis at the VR-1 training reactor

    Science.gov (United States)

    Stefanik, Milan; Rataj, Jan; Huml, Ondrej; Sklenka, Lubomir

    2017-11-01

    The VR-1 training reactor operated by the Czech Technical University in Prague is utilized mainly for education of students and training of various reactor staff; however, R&D is also carried out at the reactor. The experimental instrumentation of the reactor can be used for the irradiation experiments and neutron activation analysis. In this paper, the neutron activation analysis (NAA) is used for a study of dietary supplements containing the zinc (one of the essential trace elements for the human body). This analysis includes the dietary supplement pills of different brands; each brand is represented by several different batches of pills. All pills were irradiated together with the standard activation etalons in the vertical channel of the VR-1 reactor at the nominal power (80 W). Activated samples were investigated by the nuclear gamma-ray spectrometry technique employing the semiconductor HPGe detector. From resulting saturated activities, the amount of mineral element (Zn) in the pills was determined using the comparative NAA method. The results show clearly that the VR-1 training reactor is utilizable for neutron activation analysis experiments.

  4. Argentina: Disposal aspects of RA-1 research reactor decommissioning waste

    Energy Technology Data Exchange (ETDEWEB)

    Harriague, S; Barberis, C; Cinat, E; Grizutti, C; Scolari, H [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

    2007-12-15

    The objective of the project is to analyze disposal aspects of waste from total dismantling of Argentinean research reactors, starting with the oldest one, 48 years old RA-1. In order to estimate decommissioning waste, data was collected from files, area monitoring, measurements, sampling to measure activity and composition, operational history and tracing of operational incidents. Measurements were complemented with neutron activation calculations. Decommissioning waste for RA-1 is estimated to be 71.5 metric tons, most of it concrete (57 tons), the rest being steels, lead and reflector graphite (4.8 tons). Due to their low specific activities, no disposal problems are foreseen in the case of metals and concrete. Disposal of aluminium, steel, lead and concrete is analyzed. On the contrary, as the country has no experience in managing graphite radioactive waste, work was concentrated on that material. Stored (Wigner) energy may exist in RA-1 graphite reflectors irradiated at room temperature. Evaluation of stored energy by calorimetric methods is proposed, and its annealing by inductive heating; HEPA filters should be used to deal with gaseous activity emissions, mainly Cl-36 and C-14. Galvanic corrosion, dust explosion, ignition and oxidation can be addressed and should not become disposal problems. Care must be taken with graphite dust generation and disposal, due to wetting and flotation problems. Lessons learned from the project are presented, and the benefits of sharing international experience are stressed. (author)

  5. Subchannel analysis of 37-rod tight-lattice bundle experiments for reduced-moderation water reactor

    International Nuclear Information System (INIS)

    Nakatsuka, Toru; Tamai, Hidesada; Akimoto, Hajime

    2005-01-01

    R and D project to investigate thermal-hydraulic performance of tight-lattice fuel bundles for Reduced-Moderation Water Reactor (RMWR) started at Japan Atomic Energy Research Institute (JAERI) in collaboration with utilities, reactor vendors and universities from 2002. The RMWR realizes a high conversion ratio larger than 0.1 for sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The reactor core comprises tight-lattice fuel assemblies with gap clearance of around 1.0 mm to reduce the water volume ratio to achieve the high conversion ratio. A problem of utmost importance from a thermal-hydraulic point of view is the coolability of the tight-lattice assembly with such a small gap width. JAERI has been carrying out experimental study to investigate the system parameter effects on the thermal-hydraulic performance and to confirm the feasibility of the core. In the present study, the subchannel analysis code NASCA was applied to 37-rod tight-lattice bundle experiments. The NASCA can give good predictions of critical power for the gap width of 1.3 mm while the prediction accuracy decreases for the gap width of 1.0 mm. To improve the prediction accuracy, the code will be modified to take the effect of film thickness distribution around fuel rods on boiling transition. (author)

  6. Some considerations for assurance of reactor safety from experiences in research reactors

    International Nuclear Information System (INIS)

    Okamoto, Sunao; Nishihara, Hideaki; Shibata, Toshikazu

    1981-01-01

    For the purpose of assuring reactor safety and strengthening research in the related fields, a multi-disciplinary group was formed among university researchers, including social scientists, with a special allocation of the Grant-in-Aid from the Ministry of Education, Science and Culture. An excerpt from the first year's report (1979 -- 1980) is edited here, which contains an interpretation of Murphy's reliability engineering law, a scope of reactor diagnostic studies to be pursued at universities, and safety measures already implemented or suggested to be implemented in university research reactors. (author)

  7. Experience with automatic reactor control at EBR-II

    International Nuclear Information System (INIS)

    Lehto, W.K.; Larson, H.A.; Christensen, L.J.

    1985-01-01

    Satisfactory operation of the ACRDS has extended the capabilities of EBR-II to a transient test facility, achieving automatic transient control. Test assemblies can now be irradiated in transient conditions overlapping the slower transient capability of the TREAT reactor

  8. Research reactor decommissioning experience - concrete removal and disposal -

    International Nuclear Information System (INIS)

    Manning, Mark R.; Gardner, Frederick W.

    1990-01-01

    Removal and disposal of neutron activated concrete from biological shields is the most significant operational task associated with research reactor decommissioning. During the period of 1985 thru 1989 Chem-Nuclear Systems, Inc. was the prime contractor for complete dismantlement and decommissioning of the Northrop TRIGA Mark F, the Virginia Tech Argonaut, and the Michigan State University TRIGA Mark I Reactor Facilities. This paper discusses operational requirements, methods employed, and results of the concrete removal, packaging, transport and disposal operations for these (3) research reactor decommissioning projects. Methods employed for each are compared. Disposal of concrete above and below regulatory release limits for unrestricted use are discussed. This study concludes that activated reactor biological shield concrete can be safely removed and buried under current regulations

  9. Numerical analysis and scale experiment design of the hot water layer system of the Brazilian Multipurpose Reactor (RMB reactor)

    International Nuclear Information System (INIS)

    Schweizer, Fernando Lage Araújo

    2014-01-01

    The Brazilian Multipurpose Reactor (RMB) consists in a 30 MW open pool research reactor and its design is currently in development. The RMB is intended to produce a neutron flux applied at material irradiation for radioisotope production and materials and nuclear fuel tests. The reactor is immersed in a deep water pool needed for radiation shielding and thermal protection. A heating and purifying system is applied in research reactors with high thermal power in order to create a Hot Water Layer (HWL) on the pool top preventing that contaminated water from the reactor core neighboring reaches its surface reducing the room radiation dose rate. This dissertation presents a study of the HWL behavior during the reactor operation first hours where perturbations due to the cooling system and pool heating induce a mixing flow in the HWL reducing its protection. Numerical simulations using the CFD code CFX 14.0 have been performed for theoretical dose rate estimation during reactor operation, for a 1/10 scaled down model using dimensional analysis and mesh testing as an initial verification of the commercial code application. Equipment and sensor needed for an experimental bench project were defined by the CFD numerical simulation. (author)

  10. Experience and prospects of WWER-1000 reactor spent fuel transport

    International Nuclear Information System (INIS)

    Kondratyev, A.N.; Yershov, V.N.; Kozlov, Yu.V.; Kosarev, Yu.A.; Ilyin, Yu.V.; Pavlov, M.S.

    1989-01-01

    The paper deals with the USSR experience in shipping the commercial WWER-1000 reactor spent fuel in TK-10 and TK-13 casks. The cask designs, their basic characteristics and the WWER-1000 spent fuel features are described. An example of calculational/experimental approach in the design of a basket (one of the most important components) for spent fuel assembly (SFA) accommodation in a cask is given. The main problems of future development works are presented in brief. A concept of development of nuclear power industry with the closed fuel cycle is assumed in the Soviet Union, hence the spent nuclear fuel is to be transported from NPPs to reprocessing plants. To transport WWER-1000 spent fuel, the casks of two types were developed. These are: a pilot TK-10 cask of 3t capacity in fuel; a commercial TK-13 cask of ∼6t capacity in fuel. The pilot TK-10 cask is thick-walled (360mm) cylindrical vessel manufactured of steel shells and a bottom welded to each other. The material of the body is carbon steel. There is a steel jacket on the outer side of the cask body and at 120 mm distance off the bottom. On its cylindrical part between the jacket and the body there are T-shaped circular ribs acting as shock-absorbers. The space between the jacket and the body is filled with ethylene glycol solution of 65 degree C crystallization temperature, which functions as a neutron shielding. The TK-10 cask coolant is water or air (nitrogen) at minor excess pressure resulted from FA heatup after the cask sealing

  11. Department of Reactor Technology annual progress report 1 January -31 December 1977

    International Nuclear Information System (INIS)

    1978-04-01

    The work of the Department of Reactor Technology within the following fields is described: reactor engineering, reactor operation, structural reliability, system reliability, reactor physics, fuel management, reactor accident analysis for LOCA and ECC, containment analysis, experimental heat transfer, reactor core dynamics and power plant simulators, experimental activation measurements and neutron radiography at the DR 1 reactor, underground storage of gas, solar heating and underground heat storage, wind power. (author)

  12. Predictions on an HTR coolant composition after operational experience with experimental reactors

    International Nuclear Information System (INIS)

    Nieder, R.

    1981-01-01

    Long-term operational experience of the HTR experimental reactors Dragon (1966 - 1975), Peach Bottom (1967 - 1974) and AVR (since 1967) has yielded a large number of common quantitative and qualitative results about the sources and behaviour of helium impurities in the primary circuits. Additional information has also been obtained from experiments made at the three reactors. The results at the AVR are particularly interesting because the gas outlet temperature can be varied from 770 0 C to 950 0 C when the reactor power is kept constant. Hence they can be studied according to the temperature dependence of all chemical reactions. It should be possible to apply the results from the operating measurements and experiments made at the reactors, in particular the interrelation of the impurity concentrations, to future reactors. The absolute values of these impurity concentrations are obtained first and foremost by the corresponding helium purification constants

  13. Calculation to experiment comparison of SPND signals in various nuclear reactor environments

    Energy Technology Data Exchange (ETDEWEB)

    Barbot, Loic; Radulovic, Vladimir; Fourmentel, Damien [CEA, DEN, DER, Instrumentation, Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance, (France); Snoj, Luka [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana, (Slovenia); Tarchalski, Mikolaj [National Centre for Nuclear Research, ulica Andrzeja Soltana 7, 05-400 Otwock (Swierk), (Poland); Dewynter-Marty, Veronique [CEA, DEN, DANS, DRSN, SIREN, LESCI, Saclay, F-91191 Gif sur Yvette, (France); Malouch, Fadhel [CEA, DEN, DANS, DM2S, SERMA, Saclay, F-91191 Gif sur Yvette, (France)

    2015-07-01

    In the perspective of irradiation experiments in the future Jules Horowitz Reactor (JHR), the Instrumentation Sensors and Dosimetry Laboratory of CEA Cadarache (France) is developing a numerical tool for SPND design, simulation and operation. In the frame of the SPND numerical tool qualification, dedicated experiments have been performed both in the Slovenian TRIGA Mark II reactor (JSI) and very recently in the French CEA Saclay OSIRIS reactor, as well as a test of two detectors in the core of the Polish MARIA reactor (NCBJ). A full description of experimental set-ups and neutron-gamma calculations schemes are provided in the first part of the paper. Calculation to experiment comparison of the various SPNDs in the different reactors is thoroughly described and discussed in the second part. Presented comparisons show promising final results. (authors)

  14. Experience and lessons learned in the assessment of safety justifications for experiments mounted in research reactors

    International Nuclear Information System (INIS)

    Cox, R.F.

    1990-01-01

    Some experiments in research reactors are arguably a risky undertaking due to their uncertain outcome. The justifications for such experiments require careful assessment to validate their undertaking. The public, the operators and the installation itself must be safeguarded. Assessment of the potential risk is an acquired skill but in doing so the route can be eased by learning from the lessons experience can teach. This paper, essentially for the usage of safety managers, sets out some of the issues relating to the assessment process gained from our experience over a few tens of years in the assessment of experiments. Many of the conclusions reached may appear all too obvious viewed in retrospect, but they were not necessarily clear at the time. Those organizations setting up assessment teams may find some of the conclusions of value such that their proposed management system can embrace methodologies for assessment that can avoid or lessen the impact of some of the pitfalls we have tried to identify. Failure to recognise some of these points may run the risk of delayed clearances, dilated timescales and cost overruns. It is in the hope of reducing all these penalties that we offer our experiences

  15. Operation characteristics and conditions of training reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.; Kolros, A.; Polach, S.; Sklenka, L.

    1994-01-01

    The first 3 years of operation of the VR-1 training reactor are reviewed. This period includes its physical start-up (preparation, implementation, results) and operation development as far as the current operating configuration of the reactor core. The physical start-up was commenced using a reactor core referred to as AZ A1, whose physical parameters had been verified by calculation and whose configuration was based on data tested experimentally on the SR-0 reactor at Vochov. The next operating core, labelled AZ A2, was already prepared during the test operation of the VR-1 reactor. Its configuration was such that both of the main horizontal channels, radial and tangential, could be employed. The configuration that followed, AZ A3, was an intermediate step before testing the graphite side reflector. The current reactor core, labelled AZ A3 G, was obtained by supplementing the previous core with a one-sided graphite side reflector. (Z.S.). 2 tabs., 11 figs., 2 refs

  16. Application of the Dragon reactor experiment to the safety evaluation of current HTR systems

    International Nuclear Information System (INIS)

    Ashworth, F.P.O.; Faircloth, R.L.

    1976-01-01

    An important component of the confidence required for the safety assessment of high-temperature reactors is the experimental proof of phenomena such as fission product release or core corrosion. The most convincing experiments are those carried out in a reactor. This paper outlines the scope of experiments relevant to safety which can be done in the Dragon Reactor Experiment and describes as an example the experimental campaign and the current outcome of the work on validating the predictions of caesium release and migration. (author)

  17. On disruption of reactor core of the Chernobylsk-4 reactor (retrospective analysis of experiments and facts)

    International Nuclear Information System (INIS)

    Platonov, P.A.

    2007-01-01

    Fragments of graphite blocks from the damaged Chernobyl NPP, unit 4 are studied, the results are analyzed. The temperature of the graphite blocks at the moment of accident release from the reactor is evaluated. Results of studying the fragments of fuel channel and fuel dispersion are considered. The fuel heat content at the moment of the explosion is evaluated and some conclusions are made about the character of the reactor core destruction [ru

  18. Operating Experience from Events Reported to the IAEA Incident Reporting System for Research Reactors

    International Nuclear Information System (INIS)

    2015-03-01

    Operating experience feedback is an effective mechanism in providing lessons learned from events and the associated corrective actions to prevent them, helping to improve safety at nuclear installations. The Incident Reporting System for Research Reactors (IRSRR), which is operated by the IAEA, is an important tool for international exchange of operating experience feedback for research reactors. The IRSRR reports contain information on events of safety significance with their root causes and lessons learned which help in reducing the occurrence of similar events at research reactors. To improve the effectiveness of the system, it is essential that national organizations demonstrate an appropriate interest for the timely reporting of events important to safety and share the information in the IRSRR database. At their biennial technical meetings, the IRSRR national coordinators recommended collecting the operating experience from the events reported to the IRSRR and disseminating it in an IAEA publication. This publication highlights the root causes, safety significance, lessons learned, corrective actions and the causal factors for the events reported to the IRSRR up to September 2014. The publication also contains relevant summary information on research reactor events from sources other than the IRSRR, operating experience feedback from the International Reporting System for Operating Experience considered relevant to research reactors, and a description of the elements of an operating experience programme as established by the IAEA safety standards. This publication will be of use to research reactor operating organizations, regulators and designers, and any other organizations or individuals involved in the safety of research reactors

  19. Experimental Studies on Assemblies 1 and 2 of the Fast Reactor FR-0. Part 1

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, T L; Hellstrand, E; Londen, S O; Tiren, L I

    1965-08-15

    FR0 is a fast zero power reactor built for experiments in reactor physics. It is a split table machine containing vertical fuel elements. 120 kg of U{sup 235} are available as fuel, which is fabricated into metallic plates of 20 % enrichment. The control system comprises 5 spring-loaded safety elements and 3 + 1 elements for startup operations and power control. The reactor went critical in February 1964. The first assemblies studied were made up of undiluted fuel into a cylindrical and a spherical core, respectively, surrounded by a reflector made of copper. The present report describes some experiments made on these systems. Primarily, critical mass determinations, flux distribution measurements and studies of the conversion ratio are dealt with. The measured quantities have been compared with theoretical predictions using various transport theory programmes (DSN, TDC) and cross section sets. The experimental results show that the neutron spectrum in the copper reflector is softer than predicted, but apart from this discrepancy agreement with theory has generally been obtained.

  20. Development of the design of the High Temperature Gas Cooled Reactor experiment

    International Nuclear Information System (INIS)

    Lockett, G.E.; Huddle, R.A.U.

    1960-01-01

    Early in 1956 a small team was formed at the Atomic Energy Research Establishment, Harwell, to investigate the possibilities of the High Temperature Gas Cooled (H.T.G.C.) Reactor System. Although the primary objective of this team was to carry out a feasibility study of the system as a whole, it soon became apparent that, in addition to design studies and economic surveys of power producing reactors, the most appropriate approach to such a novel system was to carry out a design study of a relatively small (10 to 20 M.W.) Reactor Experiment, together with the necessary research and development work associated with such a reactor. This work proceeded within the U.K.A.E.A. during the three following years, and it was felt that realistic design proposals could be put forward with sufficient confidence to justify the detailed design and construction of a 20 M.W. Reactor Experiment. In April 1959 responsibility for this Reactor Experiment was taken over by the O.E.E.C. High Temperature Gas Cooled Reactor Project, the DRAGON Project, at the Atomic Energy Establishment, Winfrith, Dorset. In this Paper the research, development, and design work is reviewed, and the proposals for the Reactor Experiment are summarised. (author)

  1. Present and future activities of TRIGA RC-1 Reactor

    International Nuclear Information System (INIS)

    Festinesi, A.

    1986-01-01

    A summary of reactor activities is presented and discussed. The RC-1 reactor is used by ENEA's laboratories, research institutes and national industries for different aims: research, analysis materials behaviour under neutron flux, etc. To satisfy the requests increase it is important to signalize: - the realization of a new radiochemical laboratory for radioisotopes production, to be used in a medical and/or diagnostic field in general; - the realization of a tritium handling laboratory, to study tritium solubility, release and diffusion in different material (particularly in ceramic breeder as lithium aluminate) to support Italian programs on fusion technology; - a research activity on the reactors computerized control by a console of advanced conception. The aim of this activity is the development of an ergonomic control room that could be a reference point for the planning of the power reactor control rooms

  2. Evaluation guide for the international reactor physics experiments evaluation project (IRPhEP)

    International Nuclear Information System (INIS)

    Yamaji, Akifumi

    2013-01-01

    At present, there is an urgent need to preserve integral reactor physics experimental data including separate or special effects data for nuclear energy and technology applications and the knowledge and competence contained therein. The International Reactor Physics Evaluation Project (IRPhEP) was initiated as a pilot activity in 1999 by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June of 2003. While coordination and administration of the IRPhEP takes place at an international level, each participating country is responsible for the administration, technical direction, and priorities of the project within their respective countries. This document outlines the general presentation guidelines that evaluators should follow for the description of the experiments and all relevant experimental data in order to ensure the consistency between the evaluations published in the final Handbook. Publication templates will be used to ensure this consistency and will follow the general scheme below: 1 - Experiment identification number; 2- Date; 3 - Name of experiment (Purpose of experiment, Phenomena measured and scope); 4 - Name or designation of experimental programme; 5 - Description of facility; 6 - Description of test or experiment (Experimental configuration, Core life cycle, Experimental limitations or shortcomings); 7 - Phenomena measured (Description of results and analysis, Special features and characteristics of experiment, Measurement systems/methods and uncertainties); 8 - Duplicate or complementary experiments / other related experiments; 9 - Status of completion of the evaluation; 10 - References (pointer to evaluation, archive if available, otherwise generic bibliographic reference); 11 - Authors/ organisers 12 - Material available

  3. Electrical system regulations of the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Mello, Jose Roberto de; Madi Filho, Tufic

    2013-01-01

    The IEA-R1 reactor of the Nuclear and Energy Research Institute (IPEN-CNEN/SP), is a research reactor open pool type, designed and built by the U.S. firm Babcock and Wilcox, having, as coolant and moderator, deionized light water and beryllium and graphite, as reflectors. Until about 1988, the reactor safety systems received power from only one source of energy. As an example, it may be cited the control desk that was powered only by the vital electrical system 220V, which, in case the electricity fails, is powered by the generator group: no-break 220V. In the years 1989 and 1990, a reform of the electrical system upgrading to increase the reactor power and, also, to meet the technical standards of the ABNT (Associacao Brasileira de Normas Tecnicas) was carried out. This work has the objective of showing the relationship between the electric power system and the IEA-R1 reactor security. Also, it demonstrates that, should some electrical power interruption occur, during the reactor operation, this occurrence would not start an accident event. (author)

  4. Experiences with fast breeder reactor education in laboratory and short course settings

    International Nuclear Information System (INIS)

    Waltar, A.E.

    1983-01-01

    The breeder reactor industry throughout the world has grown impressively over the last two decades. Despite the uncertainties in some national programs, breeder reactor technology is well established on a global scale. Given the magnitude of this technological undertaking, there has been surprisingly little emphasis on general breeder reactor education - either at the university or laboratory level. Many universities assume the topic too specialized for including appropriate courses in their curriculum - thus leaving students entering the breeder reactor industry to learn almost exclusively from on-the-job experience. The evaluation of four course presentations utilizing visual aids is presented

  5. Operation experience and maintenance at the TRIGA Mark II L.E.N.A. reactor

    International Nuclear Information System (INIS)

    Gngoli, F.; Berzero, A.; Lana, F.; Rosti, G.; Meloni, S.

    2008-01-01

    The TRIGA Mark II reactor of the University of Pavia was operated in the last two years on a routine basis, mostly for neutron activation analysis purposes. Moreover the reactor was completely shutdown in the first six months of this year to allow the dismantling of the NADIR experimental setup. The paper presents: - Reactor operation from July 1990 to June 1992; - Reactor users in the time period January 1990 - December 1991; - Specific activities of some radionuclides in the filling materials; - Specific activity of some radionuclides in thermal column materials. Operations related to dismantling of NADIR experimental facility are described. Finally the new thermal column configuration is presented. Starting from the end inside the reactor tank, a graphite layer (35 cm thick) was positioned, followed by a bismuth layer (10 cm thick) to reduce gamma-ray intensity. The old graphite rods were then positioned leaving in the central part, on the equatorial plane of the thermal column, a cavity whose vertical section has 40 cm width and 20 cm height. The bottom of the cavity, towards to the reactor tank, has been lined with additional layers of graphite (10 cm), bismuth (10 cm) and again graphite (1 cm). The new configuration allowed new experiments to be performed. The cavity in the central part has been created to allow the irradiation of large biological samples such as experimental animal and human livers. This is a peculiar step in a neutron capture boron therapy project to be carried out at the University of Pavia. In order to avoid an implemented 41 Ar production in the void space between shutters and the thermal column outer end, the external surface of the thermal column has been coated with boral sheets. The neutron flux profile, both thermal and epithermal, and cadmium ratio for gold are shown. The flux distribution appears to be adequate to proceed with the neutron capture boron therapy experiment. The LENA Health Physics Service has checked all phases of

  6. The Osiris reactor. Descriptive report - Volume 1 - text

    International Nuclear Information System (INIS)

    1969-05-01

    Osiris is a pool type reactor with a 70 MW thermal power. Its main purpose is to irradiate under high flows of neutrons the materials of which future nuclear power stations are made. This report proposes a description of this pool reactor. A first part describes the functional aspects and general characteristics of all installations which are in principle definitely defined (premises, irradiation and experimentation equipment, water circuits, power supply, venting, controls). The second part addresses elements which are likely to be changed, and more particularly the reactor core: fuel elements and controls (uranium and boron load in different fuel element generations, experimental locations within the core), neutron transport aspects (calculation and experiment), and thermal aspects (power generation and removal) of the pile). The third part addresses the operation: operation cycles, stops, exploitation organisation [fr

  7. Report on safety related occurrences and reactor trips July 1, 1977 - December 31, 1977

    International Nuclear Information System (INIS)

    Andermo, L.; Sundman, B.

    1974-04-01

    This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1977 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 48 safety related occurrences and 49 reactor trips have been reported to the Nuclear Power Inspectorate. Included is also one incident June 21 in Barsebaeck 2 which was not included in the last compilation of occurrences. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips have increased nearly 30% since the last period. Those occurred during power operation however, were less. More than 50% of the reactor trips happened in the shutdown condition. The fact that even small deviations from prescribed operation result in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences withou02068NRM 0000169 450

  8. Operation experience with the TRIGA reactor Wien 2004

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2004-01-01

    The TRIGA Mark-II reactor in Vienna is now in operation for more than 42 years. The average operation time is about 230 days per year with 90 % of this time at nominal power of 250 kW. The remaining 10 % operation time is used for students' training courses at low power level. Pulse operation is rather infrequent with about 5 to 10 pulses per year. The utilization of this facility is excellent, the number of students participating in practical exercises has strongly increased, and also training courses for outside groups such as the IAEA or for the 2004 Eugene Wigner Course are using the reactor, because it is the only TRIGA reactor remaining in Austria. Therefore, there is no need for decommissioning and it is intended to operate it as long as possible into the next decade. Nevertheless, in early 2004 it was decided to prepare a report on a decommissioning procedure for a typical TRIGA Mark II reactor which lists the volumes, the activity and the weight of individual materials such as concrete, aluminium, stainless steel, graphite and others which will accumulate during this process (a summary of possible activated and contaminated materials and the activity of a single TRIGA fuel element as a function of fuel type and decay time in Bq is presented). The status of the reactor (instrumentation, fuel elements, cooling circuit, ventilation system, re-inspection and maintenance program, cost/benefit) is outlined. (nevyjel)

  9. Status of prompt gamma neutron activation analysis (PGAA) at TRR-1/M1 (Thai Research Reactor-1/Modified 1)

    Energy Technology Data Exchange (ETDEWEB)

    Asvavijnijkulchai, Chanchai; Dharmavanij, Wanchai; Siangsanan, Pariwat; Ratanathongchai, Wichian; Chongkum, Somporn [Physics Division, Office of Atomic Energy for Peace, Vibhavadi Rangsit Road, Chatuchak, Bangkok (Thailand)

    1999-08-01

    The first prompt gamma activation analysis (PGAA) was designed, constructed and installed at a 6 inch diameter neutron beam port of the Thai Research Reactor-1/Modified 1 (TRR-1/M1) since 1989. Beam characteristic were made by Gd foil irradiation, X-ray film exposing and densitometry scanning consequently. The thermal neutron flux at sample position was measured by Au foil activation, and was about 1 x 10{sup 7} n.cm{sup 2}.sec{sup -1} at 700 kW operating power. The experiments have been conducted successfully. In 1998, the PGAA facility has been developed for the reactor operating power at 1.2 MW. The new PGAA system, e.g., beam shutter, gamma collimator and biological shields have been designed to reduce the leakage of neutrons and gamma radiation to the acceptance levels in accordance with the International Commission on Radiation Protection Publication 60 (ICRP 60). The construction and installation will be completed in April 1999. (author)

  10. Recent performance experience with US light water reactor self-actuating safety and relief valves

    Energy Technology Data Exchange (ETDEWEB)

    Hammer, C.G.

    1996-12-01

    Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

  11. Research program plan: reactor vessels. Volume 1

    International Nuclear Information System (INIS)

    Vagins, M.; Taboada, A.

    1985-07-01

    The ability of the licensing staff of the NRC to make decisions concerning the present and continuing safety of nuclear reactor pressure vessels under both normal and abnormal operating conditions is dependent upon the existence of verified analysis methods and a solid background of applicable experimental data. It is the role of this program to provide both the analytical methods and the experimental data needed. Specifically, this program develops fracture mechanics analysis methods and design criteria for predicting the stress levels and flaw sizes required for crack initiation, propagation, and arrest in LWR pressure vessels under all known and postulated operations conditions. To do this, not only must the methods be developed but they must be experimentally validated. Further, the materials data necessary for input to these analytical methods must be developed. Thus, in addition to methods development and large scale experimental verification this program also develops data to show that slow-load fracture toughness, rapid-load fracture toughness, and crack arrest toughness obtained from small laboratory specimens are truly representative of the toughness characteristics of the material behavior in pressure vessels in both the unirradiated and the irradiated conditions

  12. Diamond Ordinance Radiation Facility (DORF) reactor operating experiences

    International Nuclear Information System (INIS)

    Gieseler, Walter

    1970-01-01

    The Diamond Ordnance Radiation Facility Mark F Reactor is described and some of the problems encountered with its operation are discussed. In a period from reactor startup in September 1961 to June 1964, when the aluminum-clad core was changed to a stainless-steel clad core, a total of 30 fuel elements were removed from reactor service because of excessive growth. One leaking fuel element was detected during the lifetime of the aluminum- clad core. In June 1964, the core was changed to the stainless-steel-clad high hydride fuel elements. Since the installation of the stainless-steel-clad fuel element core, there has been a gradual decline of excess reactivity. Various theories were discussed as the cause but the investigations have resulted in no definitive conclusion that could account for the total reactivity loss

  13. Determination of reactor parameters by single rod experiments

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N; Zdravkovic, Z; Ivkovic, M [Department of Reactor Physics and Dynamics, Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia)

    1969-07-01

    A method is developed for the experimental determination of reactor parameters by using an isolated fuel element. The method is based on the consideration of the fuel element as the source and sink of neutrons when placed in a constant neutron field. By measuring the perturbation of the original field produced by insertion of the test fuel element it was possible to determine the fuel element parameters defined by the heterogeneous reactor theory of Feinberg and Galanin as thermal neutron absorption constant {gamma}, and neutron multiplication constant {eta}. Statistical error for one series of measurement amount to 2% in the values of {eta} and {gamma}. The developed method was intended for the analysis of the nuclear characteristics of the fuel element in the stage of its construction and development for a given reactor system. (author)

  14. System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

    International Nuclear Information System (INIS)

    Moiseyev, A.V.

    2008-01-01

    There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k eff , control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

  15. System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Moiseyev, A.V. [SSC RF - IPPE, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)

    2008-07-01

    There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k{sub eff}, control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

  16. Feedback from Westinghouse experience on segmentation of reactor vessel internals - 59013

    International Nuclear Information System (INIS)

    Kreitman, Paul J.; Boucau, Joseph; Segerud, Per; Fallstroem, Stefan

    2012-01-01

    With more than 25 years of experience in the development of reactor vessel internals segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. Building on tooling concepts and cutting methodologies developed decades ago for the successful removal of nuclear fuel from the damaged Three Mile Island Unit 2 reactor (TMI-2), Westinghouse has continuously improved its approach to internals segmentation and packaging by incorporating lessons learned and best practices into each successive project. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive water-jet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Westinghouse has applied its technology to all types of reactors covering Pressurized Water Reactors (PWR's), Boiling Water Reactors (BWR's), Gas Cooled Reactors (GCR's) and sodium reactors. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since space is almost always a limiting factor it is therefore important to plan and optimize the available room in the segmentation areas. The choice of the optimum cutting technology is important for a successful project implementation and depends on some specific constraints like disposal costs, project schedule, available areas or safety. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. Westinghouse has also developed a variety of special handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a

  17. System Requirements Document for the Molten Salt Reactor Experiment 233U conversion system

    International Nuclear Information System (INIS)

    Aigner, R.D.

    2000-01-01

    The purpose of the conversion process is to convert the 233 U fluoride compounds that are being extracted from the Molten Salt Reactor Experiment (MSRE) equipment to a stable oxide for long-term storage at Bldg. 3019

  18. Determination of reactor parameters by single rod experiments

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N; Zdravkovic, Z; Ivkovic, M; Sotic, O [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1968-10-15

    The objective of this work was to determine experimentally fuel element parameters using an isolated fuel element of arbitrary construction and analyzing the accuracy of their results with the aim to apply them in analysis of reactor system. The approach is based on assumption of heterogeneous reactor theory, 'source-sink' theory. The obtained experimental results have shown the possibility of obtaining data for absorption or production properties of fuel element by analyzing the thermal and epithermal neutron density distributions around a single fuel rod placed in a sufficiently large thermal hole.

  19. Safety Research Experiment Facility Project. Conceptual design report. Volume V. Reactor vessel and closure

    International Nuclear Information System (INIS)

    1975-12-01

    The Prestressed Concrete Reactor Vessel (PCRV) will serve as the primary pressure retaining structure for the Safety Research Experiment Facility (SAREF) reactor. The reactor core, control rod drive room, primary heat exchangers, and gas circulators will be located in cavities within the PCRV. The orientation of these cavities, except for the control rod drive room, will be similar to the high-temperature gas-cooled reactor (HTGR) designs that are currently proposed or under design. Due to the nature of this type of structure, all biological and radiological shielding requirements are incorporated into the basic vessel design. At the midcore plane there are three radially oriented slots that will extend from the outside surface of the PCRV to the reactor core liner. These slots will accommodate each of the fuel motion monitoring systems which will be part of the observation apparatus used with the loop experiments

  20. Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

    International Nuclear Information System (INIS)

    Song, C. H.; Chung, M. K.; Park, C. K. and others

    2005-04-01

    The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved

  1. U-233 fuelled low critical mass solution reactor experiment PURNIMA II

    International Nuclear Information System (INIS)

    Srinivasan, M.; Chandramoleshwar, K.; Pasupathy, C.S.; Rasheed, K.K.; Subba Rao, K.

    1987-01-01

    A homogeneous U-233 uranyl nitrate solution fuelled BeO reflected, low critical mass reactor has been built at the Bhabha Atomic Research Centre, India. Christened PURNIMA II, the reactor was used for the study of the variation of critical mass as a function of fuel solution concentration to determine the minimum critical mass achievable for this geometry. Other experiments performed include the determination of temperature coefficient of reactivity, study of time behaviour of photoneutrons produced due to interaction between decaying U-233 fission product gammas and the beryllium reflector and reactor noise measurements. Besides being the only operational U-233 fuelled reactor at present, PURNIMA II also has the distinction of having attained the lowest critical mass of 397 g of fissile fuel for any operating reactor at the current time. The paper briefly describes the facility and gives an account of the experiments performed and results achieved. (author)

  2. Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Song, C. H.; Chung, M. K.; Park, C. K. and others

    2005-04-15

    The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved.

  3. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities

  4. Research reactors for power reactor fuel and materials testing - Studsvik's experience

    International Nuclear Information System (INIS)

    Grounes, M.

    1998-01-01

    Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

  5. Test on the reactor with the portable digital reactivity meter for physical experiment

    International Nuclear Information System (INIS)

    Huang Liyuan

    2010-01-01

    Test must be performed on the zero power reactor During the development of portable digital reactivity meter for physical experiment, in order to check its measurement function and accuracy. It describes the test facility, test core, test methods, test items and test results. The test results show that the instrument satisfy the requirements of technical specification, and satisfy the reactivity measurement in the physical experiments on reactors. (authors)

  6. RA Research reactor Annual report 1982 - Part 1, Operation, maintenance and utilization of the RA reactor

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.; Miokovic, J.

    1982-12-01

    Reactor test operation started in September 1981 at 2 MW power with 80% enriched fuel continued during 1982 according to the previous plan. The initial reactor core was made of 44 fuel channel each containing 10 fuel slugs. The first half of 1982 was used for the needed measurements and analysis of operating parameters and functioning of reactor systems and equipment under operating conditions. Program concerned with the testing operation at higher power levels was started in the second half of this year. It was found that the inherent excess reactivity and control rod worths ensure safe operation according to the IAEA safety standards. Excess reactivity is high enough to enable higher power level of 4.7 MW during 4 monthly cycles each lasting 15-20 days. Favourable conditions for cooling exist for the initial core configuration. Effects of poisoning at startup on the reactivity and power density distribution were measured as well as initial spatial distribution of the neutron flux which was 3,9 10 13 cm -2 s -1 at 2 MW power. Modification of the calibration coefficient in the system for automated power level control was determined. All the results show that all the safety criteria and limitations concerned with fuel utilization are fulfilled if reactor power would be 4.7 MW. Additional testing operation at 3, 4, and 4.7 MW power levels will be needed after obtaining the licence for operating at nominal power. Transition from the initial core with 44 fuel channels to the equilibrium lattice configuration with 72 fuel channels each containing 10 fuel slugs, would be done gradually. Reactor was not operated in September because of the secondary coolant pipes were exchanged between Danube and the horizontal sedimentary. Control and maintenance of the reactor equipment was done regularly and efficiently dependent on the availability of the spare parts. Difficulties in maintenance of the reactor instrumentation were caused by unavailability of the outdated spare parts

  7. Basic considerations for the safety analysis report of the Greek Research Reactor-1 (GRR-1)

    International Nuclear Information System (INIS)

    Anoussis, J.N.; Chrysochoides, N.G.; Papastergiou, C.N.

    1980-09-01

    The basic considerations upon which the new revised Safety Analysis Report (SAR) for the GRR-1 will be based are presented. The format and the content the SAR will follow are given. A number of credible and less credible accidents is briefly analysed on the basis of present knowledge and experience for similar reactors, as well as the experience gained in the last 10 years of the GRR-1 operation at 5 MW. The accident caused by partial blockage of the cooling flow is considered to be the Maximum Credible Accident (MCA) for the GRR-1. The MCA is analysed and its radiological impact to the environment is estimated using conservative assumptions. (T.A.)

  8. Experience of joint reactor laboratory course with KUCA

    International Nuclear Information System (INIS)

    Nishina, Kojiro

    1982-01-01

    A description is given of a joint reactor laboratory course of graduate level, which is offered every summer since 1975 by nine associated japanese universities with the use of Kyoto University Critical Assembly (KUCA). A total of 315 students have taken the course in the last seven years. The course has been institutionalized with the background that it is extremely difficult for any single university in this country to have her own research or training reactor. By their effort the united faculty team of the course have succeeded in giving an effective, unique one-week course, taking advantage of their collaboration. By the scrutiny of student's responses to the course, one norices that a reactor is distinctively different from ordinary educational apparatus, in that the students do not play a main part in the adjustment of the apparatus itself. The beginners therefore tend to feel the reactor as a remote existence. This difficulty must be circumvented if an effective educational process is to be designed. (author)

  9. Worldwide experience with light water reactor fuel - a review

    International Nuclear Information System (INIS)

    Strasser, A.A.

    1986-01-01

    Continued attention to fuel performance has over the years improved fuel reliability and reduced fuel related failures. But further improvements can still be made by increased attention to reactor operating and maintenance methods, as well as to quality control during fuel fabrication. (author)

  10. Operating experiences of reactor shutdown system at MAPS

    International Nuclear Information System (INIS)

    Kotteeswaran, T.J.; Subramani, V.A.; Hariharan, K.

    1997-01-01

    The reactors in Madras Atomic Power Station (MAPS), Kalpakkam are Pressurised Heavy Water Reactors (PHWR) similar to RAPS, Kota. The moderator heavy water is pumped into the calandria from dump tank to make the reactor critical. Later with the calandria level held constant at 92% FT, the further power changes are being done with the movement of adjuster rods. The moderator is held in calandria by means of helium gas pressure differential between top of calandria and dump tank located below. The shutdown of the reactor is effected by dumping the moderator water to dump tank by fast equalizing of helium gas pressure. In the revised mode of operation of moderator circuit after the moderator inlet manifold failure, the dump timing was observed to be more compared to the normal value. This was investigated and observed to be due to accumulation of D 2 O in the gas space above dump valves, which was affecting the helium equalizing flow. Also some of Indicating Alarm Meters (IAM) in protective system initiating the trip signals have failed in the unsafe mode. They have been modified to avoid the recurrence of the failures. (author)

  11. Operating experiences with 1 MW steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Sano, A; Kanamori, A; Tsuchiya, T

    1975-07-01

    1 MW steam generator, which was planned as the first stage of steam generator development in Power Reactor and Nuclear Fuel Corp. (PNC) in Japan, is a single-unit, once-through, integrated shell and tube type with multi-helical coil tubes. It was completed in Oarai Engineering Center of PNC in March of 1971, and the various performance tests were carried out up to April, 1972. After the dismantle of the steam generator for structural inspection and material test, it was restored with some improvements. In this second 1 MW steam generator, small leak occurred twice during normal operation. After repairing the failure, the same kind of performance tests as the first steam generator were conducted in order to verify the thermal insulation effect of argon gas in downcomer zone from March to June, 1974. In this paper the above operating experiences were presented including the outline of some performance test results. (author)

  12. Burnup measurements at the RECH-1 research reactor

    International Nuclear Information System (INIS)

    Henriquez, C.; Navarro, G.; Pereda, C.; Torres, H.; Pena, L.; Klein, J.; Calderon, D.; Kestelman, A.J.

    2002-01-01

    The Chilean Nuclear Energy Commission has decided to produce LEU fuel elements for the RECH-1 research reactor. During December 1998, the Fuel Fabrication Plant delivered the first four fuel elements, called leaders, to the RECH-1 reactor. The set was introduced into the reactor's core, following the normal routine, but performing a special follow-up on their behavior inside and outside the core. In order to measure the burn-up of the leader fuel elements, it was decided to develop a burn-up measurements system to be installed into the RECH-1 reactor pool, and to decline the use of a similar system, which operates in a hot cell. The main reason to build this facility was to have the capability to measure the burn-up of fuel elements without waiting for long decay period. This paper gives a brief description of the facility to measure the burn-up of spent fuel elements installed into the reactor pool, showing the preliminary obtained spectra and briefly discussing them. (author)

  13. Bench-mark experiments to study the neutron distribution in a heterogeneous reactor shielding

    International Nuclear Information System (INIS)

    Bolyatko, V.V.; Vyrskij, M.Yu.; Mashkovich, V.P.; Nagaev, R.Kh.; Prit'mov, A.P.; Sakharov, V.K.; Troshin, V.S.; Tikhonov, E.G.

    1981-01-01

    The bench-mark experiments performed at the B-2 facility of the BR-10 reactor to investigate the spatial and energy neutron distributions are described. The experimental facility includes the neutron beam channel with a slide, a mo shielding composition investigated consisted of sequential layers of steel (1KH18N9T) and graphite slabs. The neutron spectra were measured by activation method, a set of treshold and resonance detectors having been used. The detectors made it possible to obtain the absolute neutron spectra in the 1.4 eV-10 MeV range. The comparison of calculations with the results of the bench-mark experiments made it possible to prove the neutron transport calculational model realized in the ROZ-9 and ARAMAKO-2F computer codes and evaluate the validity of the ARAMAKO constants for the class of shielding compositions in question [ru

  14. Flooding Experiments and Modeling for Improved Reactor Safety

    International Nuclear Information System (INIS)

    Solmos, M.; Hogan, K.J.; VIerow, K.

    2008-01-01

    Countercurrent two-phase flow and 'flooding' phenomena in light water reactor systems are being investigated experimentally and analytically to improve reactor safety of current and future reactors. The aspects that will be better clarified are the effects of condensation and tube inclination on flooding in large diameter tubes. The current project aims to improve the level of understanding of flooding mechanisms and to develop an analysis model for more accurate evaluations of flooding in the pressurizer surge line of a Pressurized Water Reactor (PWR). Interest in flooding has recently increased because Countercurrent Flow Limitation (CCFL) in the AP600 pressurizer surge line can affect the vessel refill rate following a small break LOCA and because analysis of hypothetical severe accidents with the current flooding models in reactor safety codes shows that these models represent the largest uncertainty in analysis of steam generator tube creep rupture. During a hypothetical station blackout without auxiliary feedwater recovery, should the hot leg become voided, the pressurizer liquid will drain to the hot leg and flooding may occur in the surge line. The flooding model heavily influences the pressurizer emptying rate and the potential for surge line structural failure due to overheating and creep rupture. The air-water test results in vertical tubes are presented in this paper along with a semi-empirical correlation for the onset of flooding. The unique aspects of the study include careful experimentation on large-diameter tubes and an integrated program in which air-water testing provides benchmark knowledge and visualization data from which to conduct steam-water testing

  15. Metal fire implications for advanced reactors. Part 1, literature review

    International Nuclear Information System (INIS)

    Nowlen, Steven Patrick; Radel, Ross F.; Hewson, John C.; Olivier, Tara Jean; Blanchat, Thomas K.

    2007-01-01

    Public safety and acceptance is extremely important for the nuclear power renaissance to get started. The Advanced Burner Reactor and other potential designs utilize liquid sodium as a primary coolant which provides distinct challenges to the nuclear power industry. Fire is a dominant contributor to total nuclear plant risk events for current generation nuclear power plants. Utilizing past experience to develop suitable safety systems and procedures will minimize the chance of sodium leaks and the associated consequences in the next generation. An advanced understanding of metal fire behavior in regards to the new designs will benefit both science and industry. This report presents an extensive literature review that captures past experiences, new advanced reactor designs, and the current state-of-knowledge related to liquid sodium combustion behavior

  16. Operational Experience with the TRIGA Mark II Reactor of the University of Pavia

    Energy Technology Data Exchange (ETDEWEB)

    Tigliole, A. Borio Di; Alloni, D.; Cagnazzo, M.; Coniglio, M.; Lana, F.; Losi, A.; Magrotti, G.; Manera, S.; Marchetti, F.; Pappalardo, P.; Prata, M.; Provasi, M.C.; Salvini, A.; Scian, G.; Vinciguerra, G. [University of Pavia, Laboratory of Applied Nuclear Energy (L.E.N.A), Via Aselli 41, 27100 Pavia (Italy)

    2011-07-01

    The Laboratory of Applied Nuclear Energy (LENA) is an Interdepartmental Research Centre of the University of Pavia which operates a 250 kW TRIGA Mark II Research Nuclear Reactor, a Cyclotron for the production of radioisotopes and other irradiation facilities. The reactor is in operation since 1965 and many home-made upgrading were realized in the past years in order to assure a continuous operation of the reactor for the future. The annual reactor operational time at nominal power is in the range of 300 - 400 hours depending upon the time schedule of some experiments and research activities. The reactor is mainly used for NAA activities, BNCT research, samples irradiation and training. In specific, few tens of hours of reactor operation per year are dedicated to training courses for University students and for professionals. Besides, the LENA Centre hosts every year more than one thousand high school students in visit. Lately, LENA was certified ISO 9001:2008 for the ''operation and maintenance of the reactor'' and for the ''design and delivery of the irradiation service''. Nowadays the reactor shows a good technical state and, at the moment, there are no political or economical reason to consider the reactor shut-down. (author)

  17. Reliabitity study of the accumulator system for Angra-1 reactor

    International Nuclear Information System (INIS)

    Santos Maciel, C.C.R.

    1980-01-01

    The realibility of the Accumulator System of Angra 1 reactor is studied. The fault tree techniques is use for identification and evaluation of the probability of occurrence of the possible failure modes of the system. The study has as a guide the report WASH 1400 in which the analysis of the reliability of a Tipical PWR reactor of USA. Comparisons between results obtained for Accumulator System of Angra 1 and that published in the report WASH 1400 for the Accumulator System of the Typical Reactor are done. Critiques to the methodology used in the reportd WASH 1400 and an analysis of the sensitivity of the system in relation with its components are also done. (author) [pt

  18. IRPhE/STEK, Reactor Physics Experiments from Fast-Thermal Coupled Facility

    International Nuclear Information System (INIS)

    Dietze, Klaus; Klippel, Henk Th.; Koning, Arjan; Jacqmin, Robert

    2003-01-01

    1 - Description: The STEK-experiments have been performed to check neutron data of the most important reactor materials, especially of fission product nuclides, fuel isotopes and structural materials. The measured central reactivity worths (CRW) of small samples were compared with calculated values. These C/E-ratios have been used then for data corrections or in adjustment procedures. The reactors STEK (ECN Petten/ Netherlands) was a fast-thermal coupled facility of zero power. The annular thermal drivers were filled by fuel assemblies and moderated by water. The inner insertion lattices were loaded with pellets of fuel and other materials producing the fast neutron flux. The characteristics of the neutron and adjoint spectra were obtained by special arrangements of these pellets in unit cells. In this way, a hard or soft neutron spectrum or a special energy behavior of the adjoint function could be reached. The samples were moved by means of tubes to the central position (pile-oscillation technique). The original information about the facility and measurements is compiled in RCN-209, ECN-10 The 5 STEK configurations cover a broad energy range due to their increasing softness. The experiments are very valuable because of the extensive program of sample reactivity measurements with many fission product nuclides important in reactor burn-up calculations. At first, analyses of the experiments have been performed in Petten. Newer analyses were done later in Cadarache / CEA France using the European scheme for reactor calculation JEF-2.2 / ECCO / ERANOS (see Note Techniques and JEF/DOC-746). Furthermore, re-analyses were performed in O-arai / JNC Japan with the JNC standard route JENDL-3.2 / SLAROM / CITATION / PERKY. Results obtained with both code systems and different data evaluations (JEF-2.2 and JENDL-3.2) are compared in JEF/DOC-861. It contains the following documents: 31 Reports, 2 publications, 5 JEF documents, 4 conferences. 2 - Related or auxiliary programs

  19. Modeling and experiments on tritium permeation in fusion reactor blankets

    Science.gov (United States)

    Holland, D. F.; Longhurst, G. R.

    The determination of tritium loss from helium-cooled fusion breeding blankets are discussed. The issues are: (1) applicability of present models to permeation at low tritium pressures; (2) effectiveness of oxide layers in reducing permeation; (3) effectiveness of hydrogen addition as a means to lower tritium permeation; and (4) effectiveness of conversion to tritiated water and subsequent trapping to reduce permeation. Theoretical models applicable to these issues are discussed, and results of experiments in two areas are presented; permeation of mixtures of hydrogen isotopes and conversion to tritiated water.

  20. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  1. Department of Reactor Technology: annual progress report 1 January - 31 December 1976

    International Nuclear Information System (INIS)

    1977-06-01

    The work of the Department of Reactor Technology within the following fields is described: reactor engineering, structural reliability, system reliability, radiation fiels in nuclear power plants, reactor physics, fuel management, fission product decay analysis, steady-state thermo-hydraulics, reactor accident analysis for LOCA and ECC, containment analysis, experimental heat transfer, reactor core dynamics and power plant simulators, control rod ejection accident analysis, economic studies for power plants, experimental activation measurements and neutron radiography at the DR 1 reactor. (author)

  2. Performance demonstration experience for reactor pressure vessel shell ultrasonic testing

    International Nuclear Information System (INIS)

    Zado, V.

    1998-01-01

    The most ultrasonic testing techniques used by many vendors for pressurized water reactor (PWR) examinations were based on American Society of Mechanical Engineers 'Boiler and Pressurized Vessel Code' (ASME B and PV Code) Sections XI and V. The Addenda of ASME B and PV Code Section XI, Edition 1989 introduced Appendix VIII - 'Performance Demonstration for Ultrasonic Examination Systems'. In an effort to increase confidence in performance of ultrasonic testing of the operating nuclear power plants in United States, the ultrasonic testing performance demonstration examination of reactor vessel welds is performed in accordance with Performance Demonstration Initiative (PDI) program which is based on ASME Code Section XI, Appendix VIII requirements. This article provides information regarding extensive qualification preparation works performed prior EPRI guided performance demonstration exam of reactor vessel shell welds accomplished in January 1997 for the scope of Appendix VIII, Supplements IV and VI. Additionally, an overview of the procedures based on requirements of ASME Code Section XI and V in comparison to procedure prepared for Appendix VIII examination is given and discussed. The samples of ultrasonic signals obtained from artificial flaws implanted in vessel material are presented and results of ultrasonic testing are compared to actual flaw sizes. (author)

  3. Operation experience with the TRIGA reactor of Pavia

    International Nuclear Information System (INIS)

    Lana, F.; Marchetti, F.; Losi, A.; Orvini, E.; Borio, A.; Salvini, A.

    2002-01-01

    Operational data for the reactor for the period 2000-2002 are presented as well as an account for the irradiations, irradiated samples and reactor time requests for different applications and different users. The ventilation system has been replaced in 2001 with a new system characterised by one way through air treatments by a double stock of filters and air release through seven absolute filters (EPA 99.99%) and expulsion engine powered by an inverter. The inverter is automatically managed by a PC Honeywell in order to have over 50 Pa of depression. There is also an emergency expulsion of the air through active carbon filters. The new ventilation parameters are presented and compared to the previous values (before 2001). An account for the fuel element in the core and spent fuel elements is given. During the refuelling six new SST cladding elements have been placed in the reactor core. Configuration fuel elements have been rearranged in order to have Ring B,C, and D fill with all SST cladding elements. For the cooling system every valve has been substituted with a new one. A new cooling system display has been assembled. Pressure and flux sensors have been placed on the primary circuit

  4. Reactor power cutback system test experience at YGN 4

    International Nuclear Information System (INIS)

    Chi, Sung Goo; Kim, Se Chang; Seo, Jong Tae; Eom, Young Meen; Wook, Jeong Dae; Choi, Young Boo

    1995-01-01

    YGN 3 and 4 are the nuclear power plants having System 80 characteristics with a rated thermal output of 2815 MWth and a nominal net electrical output of 1040 MWe. YGN 3 achieved commercial operation on March 31, 1995 and YGN 4 completed Power Ascension Test (PAT) at 20%, 50%, 80% and 100% power by September 23, 1995. YGN 3 and 4 design incorporates the Reactor POwer Cutback System (RPCS) which reduces plant trips caused by Loss of Load (LOL)/ Turbine Trip and Loss of One Main Feedwater Pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety systems. The RPCS is designed to rapidly reduce reactor power by dropping preselected Control Element Assemblies (CEAs) while other NSSS control systems maintain process parameters within acceptable ranges. Extensive RPCS related tests performed during the initial startup of YGN 4 demonstrated that the RPCS can maintain the reactor on-line without opening primary or secondary safety valves and without actuating the Engineered Safety Features Actuation System (ESFAS). It is expected that use of the RPCS at YGN will increase the overall availability of the units and reduce the number of challenges to plant safety systems

  5. Probabilistic study of LOFA in ETRR-1 reactor. Vol. 4

    Energy Technology Data Exchange (ETDEWEB)

    El-Messeiry, A M [National Center for Nuclear Safety and Radiation Control, Atomic Energy Authority, Cairo (Egypt)

    1996-03-01

    In evaluating the safety of a research reactor an analysis of reactor to a wide range of postulated initiating events must be carried out, that could lead to anticipated operational occurrences or accident conditions. These disturbances include decrease in heat removal by the reactor coolant system which may be due to loss of coolant flow (LOFA) or loss of coolant heat sink. LOFA is considered here for this study for the tank type research reactor with a probabilistic approach applied to (ET-RR-1). The reactor is provided with engineering safety system to respond to accidents and perform mitigating functions. The possible malfunctions, Failures, operator errors leading to LOFA initiating event are presented (pipe break; valve opening; pump failure ...etc.). The basic event frequency/probability is calculated using appropriate probability model. The logic event tree model is constructed to illustrate all possible accident scenarios. This scenario combines system success and failure probabilities with the probability of postulated initiating events occurring that result in an accident sequence probability associated with a certain plant state. Fault tree technique is adopted to determine engineering safety features probabilities. The results show the possible minimal cut sets of variable order of each system failure. Accident sequences leading to core damage state, effects of component failures, operator errors, and system failure on plant states. The possible weak points in the design are presented. 14 figs., 3 tabs.

  6. Probabilistic study of LOFA in ETRR-1 reactor. Vol. 4

    International Nuclear Information System (INIS)

    El-Messeiry, A.M.

    1996-01-01

    In evaluating the safety of a research reactor an analysis of reactor to a wide range of postulated initiating events must be carried out, that could lead to anticipated operational occurrences or accident conditions. These disturbances include decrease in heat removal by the reactor coolant system which may be due to loss of coolant flow (LOFA) or loss of coolant heat sink. LOFA is considered here for this study for the tank type research reactor with a probabilistic approach applied to (ET-RR-1). The reactor is provided with engineering safety system to respond to accidents and perform mitigating functions. The possible malfunctions, Failures, operator errors leading to LOFA initiating event are presented (pipe break; valve opening; pump failure ...etc.). The basic event frequency/probability is calculated using appropriate probability model. The logic event tree model is constructed to illustrate all possible accident scenarios. This scenario combines system success and failure probabilities with the probability of postulated initiating events occurring that result in an accident sequence probability associated with a certain plant state. Fault tree technique is adopted to determine engineering safety features probabilities. The results show the possible minimal cut sets of variable order of each system failure. Accident sequences leading to core damage state, effects of component failures, operator errors, and system failure on plant states. The possible weak points in the design are presented. 14 figs., 3 tabs

  7. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  8. The future of the IPR-R1 TRIGA MARK I reactor after 48 years operation

    International Nuclear Information System (INIS)

    Maretti, Fausto Junior; Sette Camara, Luiz Otavio I.; Oliveira, Paulo Fernando

    2008-01-01

    The TRIGA Mark I IPR-R1 Reactor operates in the Nuclear Technology Development Center/ Brazilian Committion for Nuclear Energy (CDTN/CNEN), originally Institute of Radioactive Researches, in Belo Horizonte, Minas Gerais, since November 6, 1960. Initially it operated for isotope production for different uses, being later used in wide scale for another purposes as analyses for activation with neutrons and training of nuclear power plants operators. Dozens of degree theses were also developed with the use of the reactor. Along the years, several improvements were introduced in the reactor and its auxiliary systems, with the purpose to provide better use of the facilities and with the objective to increase the safety in the operation. The reactor is ready right now to operate at 250 kW, and for sure the nuclear applications programmed will be improved. The Operation Manual and the Safety Analysis report were already modified, as well as the Emergency Plan and the relative procedures to the same. After the tests at the end of 2008, the reactor will already be operating in the new power. This work presents a description of the several accomplishments of the last years and comments about the possibility of new uses for the reactor in the several areas of nuclear applications and some of the experiments and tests results during the upgrading program. (authors)

  9. Expanding the storage capability at ET-RR-1 research reactor at Inshass

    International Nuclear Information System (INIS)

    Sultan, Mariy M.; Khattab, M.

    1999-01-01

    Storing of spent fuel from Test Reactor in developing countries has become a big dilemma for the following reasons: The transportation of spent fuel is very expensive; There are no reprocessing plants in most developing countries; The expanding of existing storage facilities in reactor building require experience that most of developing countries lack; Some political motivations from Nuclear Developed countries intervene which makes the transportation procedures and logistics to those countries difficult. This paper gives the conceptual design of a new spent fuel storage now under construction at Inshass research reactor (ET-RR-1). The location of the new storage facility is chosen to be within the premises of the reactor facility so that both reactor and the new storage are one Material Balance Area. The paper also proposes some ideas that can enhance the transportation and storage of spent fuel of test reactors, such as: Intensifying the role of IAEA in helping countries to get rid of the spent fuel; The initiation of regional spent fuel storage facilities in some developing countries. (author)

  10. Steam--water mixing in nuclear reactor safety loss-of-coolant experiments

    International Nuclear Information System (INIS)

    Naff, S.A.; Schwarz, W.F.

    1978-01-01

    Computer models used to predict the response of reactors to hypothesized accidents necessarily incorporate approximating assumptions. To verify the models by comparing predicted and measured responses in test facilities, these assumptions must be confirmed to be realistic. Recent experiments in facilities capable of repeatedly duplicating the transient behavior of a pressurized water reactor undergoing a pipe rupture show that the assumption of complete water-steam mixing during the transient results in the predicted decompression being faster than that observed. Water reactor safety studies currently in progress include programs aimed at the verification of computer models or ''codes'' used to predict reactor system responses to various hypothesized accidents. The approach is to compare code predictions of transients with the actual test transients in experimental facilities. The purpose of this paper is to explain an important instance in which predictions and data are not in complete agreement and to indicate the significance to water reactor safety studies

  11. Life extension of the St. Lucie unit 1 reactor vessel

    International Nuclear Information System (INIS)

    Rowan, G.A.; Sun, J.B.; Mott, S.L.

    1991-01-01

    In late 1989, Florida Power and Light Company (FP and L) established the policy that St. Lucie unit 1 should not be prevented from achieving a 60-yr operating life by reactor vessel embrittlement. A 60-yr operating life means that the plant would be allowed to operate until the year 2036, which is 20 years beyond the current license expiration date of 2016. Since modifications to the reactor vessel and its components are projected to be expensive, the desire of FP and L management was to achieve this lifetime extension through the use of fuel management and proven technology. The following limitations were placed on any acceptable method for achieving this lifetime extension capability: low fuel cycle cost; low impact on safety parameters; very little or no operations impact; and use of normal reactor materials. A task team was formed along with the Advanced Nuclear Fuels Company (ANF) to develop a vessel-life extension program

  12. Modernization of Safety and Control Instrumentation of the IEA-R1 Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    De Carvalho, P.V., E-mail: paulov@ien.gov.br [Institute of Nuclear Engineering (IEN), National Nuclear Energy Commission (CNEN), Rio de Janeiro (Brazil)

    2014-08-15

    The research reactor IEA-R1 located in the Institute of Energy and Nuclear Research (IPEN), São Paulo, Brazil, obtained its first criticality on 16 September 1957 and since then has served the scientific and medical community in the performance of experiments in applied nuclear physics, as well as the provision of radioisotopes for production of radiopharmaceuticals. The reactor produces radioisotopes {sup 82}Br and {sup 41}Ar for special processes in industrial inspection and {sup 192}Ir and {sup 198}Au as sources of radiation used in brachytherapy, {sup 153}Sm for pain relief in patients with bone metastasis, and calibrated sources of {sup 133}Ba, {sup 137}Cs, {sup 57}Co, {sup 60}Co, {sup 241}Am and {sup 152}Eu used in medical clinics and hospitals practicing nuclear medicine and research laboratories. Services are offered in regular non-destructive testing by neutron radiography, neutron irradiation of silicon for phosphorous doping and other various irradiations with neutrons. The reactor is responsible for producing approximately 70% of radiopharmaceutical {sup 131}I used in Brazil, which saves about US$ 800 000 annually for the country. After more than 50 years of use, most of its equipment and systems have been modernized, and recently the reactor power was increased to 5 MW in order to enhance radioisotope production capability. However, the control room and nuclear instrumentation system used for reactor safety have operated more than 30 years and require constant maintenance. Many equipment and electronic components are obsolete, and replacements are not available in the market. The modernization of the nuclear safety and control instrumentation systems of IEA-R1 is being carried out with consideration for the internationally recognized criteria for safety and reliable reactor operations and the latest developments in nuclear electronic technology. The project for the new reactor instrumentation system specifies three wide range neutron monitoring

  13. The review of the reactor physics experiments carried out on the LR-0 research reactor NRI Rez plc for reactors of the VVER type

    International Nuclear Information System (INIS)

    Hudec, Frantisek; Jansky, Bohumil; Juricek, Vlastimil; Mikus, Jan; Novak, Evzen; Osmera, Bohumil; Posta, Severin; Rypar, Vojtech; Svadlenkova, Marie

    2010-01-01

    LR-0 is an experimental zero power reactor mainly used for the determination of the neutron-physical characteristics of WWER and PWR type reactor lattices and shielding with UO2 or MOX fuel. Its major assets include capability to design and operate multizone cores, i.e. substituted cores, with an inner inserted part in hexagonal or square geometry (driven by LR-0 standard assemblies); Standard and special supporting plates for mock-up experiments; special supporting plates, which enables the triangular symmetrical assembly arrangement with an arbitrary pitch; Modeling neutron field parameters of power reactors; Wide range benchmarking possibilities, with high reproducibility of the benchmark design parameters; Wide range of measurement techniques including equipment and experienced personal; Flexible rearrangements of the core. The main experiments included: Pin wise flux distribution measurements; VVER-440 and VVER-1000 mock-ups; compact spent fuel storage; space kinetics experiment; core parameters experimental determination; experiment with new design fuel assembly; WWER-440 control assembly influence; and burnable absorber study. International research projects are also described. (P.A.)

  14. Power transients of Ghana research reactor-1 using PARET/ANL thermal hydraulic code

    International Nuclear Information System (INIS)

    Ampomah-Amoaka, E.; Akaho, E.H.K.; Anim-Sampong, S.; Nyarko, B.J.B.

    2010-01-01

    PARET/ANL(Version 7.3 of 2007) thermal-hydraulic code was used to perform transient analysis of the Ghana Research Reactor-1.The reactivities inserted were 2.1mk and 4mk.The peak power of 5.81kW was obtained for 2.1 mk insertion whereas the peak power for 4mk insertion of reactivity was 92.32kW.These results compare closely with experiments and theoretical studies conducted previously.

  15. Simulation of the Phebus FPT1 experiment

    International Nuclear Information System (INIS)

    Amador G, R.; Nunez C, A.; Angel M, E. Del

    2003-01-01

    The present work describes the pattern of the denominated installation Phebus developed and used by the National Commission of Nuclear Security and Safeguards for their participation in the International Standard Problem ISP-46, organized by the Nuclear Energy Agency (NEA). The exercise consisted on the simulation of the denominated experiment Phebus FPT1 carried out in the experimental installation Phebus located in the Institut de Protection et de Surete Nucleaire of France. The experiment Phebus FP1 had as objective to evaluate the capacity of different computer codes to model in integral form the physical processes that are carried out during a severe accident in a pressurized water reactor (PWR), from the degradation of the core until the late stage with the formation of a pool of fused material, hydrogen production, liberation and transport of fission products, phenomena in the contention and chemistry of the iodine. The CNSNS uses the version bi of the SCDAPSIM code developed by the company Innovative Software Systems to simulate the International Standard Problem 46. The obtained results showed that the code is able to predict the thermohydraulic part of the experiment, however the same thing doesn't happen to the parameters related with the one fused of the fuel. (Author)

  16. LOCA simulation in the NRU reactor: materials test-1

    International Nuclear Information System (INIS)

    Russcher, G.E.; Marshall, R.K.; Hesson, G.M.; Wildung, N.J.; Rausch, W.N.

    1981-10-01

    A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This second experiment of the program produced peak fuel cladding temperatures of 1148K (1607 0 F) and resulted in six ruptured fuel rods. Test data and initial results from the experiment are presented here in the form of photographs and graphical summaries. These results are also compared with the preceding prototypic thermal-hydraulic test results and with computer model test predictions

  17. Irradiation experiments and materials testing capabilities in High Flux Reactor in Petten

    International Nuclear Information System (INIS)

    Luzginova, N.; Blagoeva, D.; Hegeman, H.; Van der Laan, J.

    2011-01-01

    The text of publication follows: The High Flux Reactor (HFR) in Petten is a powerful multi-purpose research and materials testing reactor operating for about 280 Full Power Days per year. In combination with hot cells facilities, HFR provides irradiation and post-irradiation examination services requested by nuclear energy research and development programs, as well as by industry and research organizations. Using a variety of the custom developed irradiation devices and a large experience in executing irradiation experiments, the HFR is suitable for fuel, materials and components testing for different reactor types. Irradiation experiments carried out at the HFR are mainly focused on the understanding of the irradiation effects on materials; and providing databases for irradiation behavior of materials to feed into safety cases. The irradiation experiments and materials testing at the HFR include the following issues. First, materials irradiation to support the nuclear plant life extensions, for instance, characterization of the reactor pressure vessel stainless steel claddings to insure structural integrity of the vessel, as well as irradiation of the weld material coupons to neutron fluence levels that are representative for Light Water Reactors (LWR) internals applications. Secondly, development and qualification of the structural materials for next generation nuclear fission reactors as well as thermo-nuclear fusion machines. The main areas of interest are in both conventional stainless steel and advanced reduced activation steels and special alloys such as Ni-base alloys. For instance safety-relevant aspects of High Temperature Reactors (HTR) such as the integrity of fuel and structural materials with increasing neutron fluence at typical HTR operating conditions has been recently assessed. Thirdly, support of the fuel safety through several fuel irradiation experiments including testing of pre-irradiated LWR fuel rods containing UO 2 or MOX fuel. Fourthly

  18. Report on safety related occurrences and reactor trips July 1, 1976-December 31, 1976

    International Nuclear Information System (INIS)

    Andermo, L.

    1977-04-01

    This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1, 1976 to December 31, 1976 inclusive. The facilities involved are Oskarshamn 1 and 2, Ringhals 1 and 2 and Barsebaeck 1. During this period of the 6 months 37 safety related occurrences and 34 reactor trips have been reported to the Nuclear Power Inspectorate. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The fact that even small deviations from prescribed operation results in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The number of reactor trips are almost as low as during the last period, which is a drastic reduction compared to earlier time periods. The greatest outages are caused by occurrences without safety significance.(author)

  19. Description of the french graphite reactor and of the experiments performed in 1956; Presentation du premier reacteur a graphite francais et des experiences effectuees en 1956

    Energy Technology Data Exchange (ETDEWEB)

    Bussac, J; Leduc, C; Zaleski, C P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1957-07-01

    This paper is an introduction to the experiments performed on the G1 reactor, experiments fully described in the papers following (670 'B to P'). The main results are given together with some comments. The neutronic parameters of the core, a description of the most important structures, and a few words of the tests leading to normal operation of the reactor under load complete our survey. (author) [French] Ce rapport presente les experiences qui furent faites sur le reacteur G1 et dont la description en detail fait l'objet des rapports suivants (670 'B a P'). Les principaux resultats sont fournis ici et commentes. On trouvera en outre les caracteristiques neutroniques du coeur actif de la pile, une description des principales installations et une mention des essais qui ont conduit au fonctionnement normal du reacteur en puissance. (auteur)

  20. ATMEA and medium power reactors. The ATMEA joint venture and the ATMEA1 medium power reactor

    International Nuclear Information System (INIS)

    Mathet, Eric; Castello, Gerard

    2012-01-01

    This Power Point presentation presents the ATMEA company (a joint venture of Areva and Mitsubishi), the main features of its medium power reactor (ATMEA1) and its building arrangement, indicates the general safety objectives. It outlines the features of its robust design which aim at protecting, cooling down and containing. It indicates the regulatory and safety frameworks, comments the review of the safety options by the ASN and the results of this assessment

  1. On the Evaluation of Pebble Bead Reactor Critical Experiments Using the Pebbed Code

    International Nuclear Information System (INIS)

    Gougar, Hans D.; Sen, R. Sonat

    2014-01-01

    Critical experiments pose a particular but necessary challenge to validating pebble bed reactor design codes. Fuel and core heterogeneities, impurities in graphite, variable packing of pebbles, and moderately strong neutronic coupling are among the factors that inject uncertainty into the results obtained with lower fidelity core physics models. Some of these are addressed in this study. The PEBBED pebble bed reactor fuel management code under development at the Idaho National Laboratory is designed for rapid design and analysis of pebble bed high temperature reactors (PBRs). Embedded within the code are the THERMIX-KONVEK thermal fluid solver and the COMBINE-7 spectrum generation code for inline cross section homogenization. Because 1D symmetry can be found at each stage of core heterogeneity; spherical at TRISO and pebble levels, and cylindrical at the control rod and core levels, the 1-D transport capability of ANISN is assumed to be sufficient in most cases for generating flux solutions for cross section homogenization. Furthermore, it is fast enough to be executed during the analysis or the equilibrium core. Multi-group diffusion-based design codes such as PEBBED and VSOP are not expected to yield the accuracy and resolution of continuous energy Monte Carlo codes for evaluation of critical experiments. Nonetheless, if the preparation of multigroup cross sections can adequately capture the physics of the mixing of PBR fuel elements and leakage from the core, reasonable results may be obtained. In this paper, results of the application of PEBBED to two critical experiments (HTR Proteus and HTR-10) and associated computational models are presented. The embedded 1-D transport solver is shown to capture the double heterogeneity of the pebble fuel in unit cell calculations. Eigenvalue calculations of a whole core are more challenging, particularly if the boron concentration is uncertain. The sensitivity of major safety parameters to variations in modeling

  2. Plutonium assemblies in reload 1 of the Dodewaard Reactor

    International Nuclear Information System (INIS)

    Bairiot, H.; Deramaix, P.; Vandenberg, C.; Leenders, L.; Mostert, P.

    1977-01-01

    Since 1963, Belgonucleaire has been developing the design of plutonium assemblies of the island type (i.e., plutonium rods inserted in the control zone of the assembly and enriched uranium rods at the periphery) for light water reactors. The application to boiling water reactors (BWRs) led to the introduction, in April 1971, of two prototype plutonium island assemblies in the Dodewaard BWR (The Netherlands): Those assemblies incorporating plutonium in 42 percent of the rods are interchangeable with standard uranium assemblies of the same reload. Their design, which had to meet these criteria, was performed using the routine order in use at Belgonucleaire; experimental checks included a mock-up configuration simulated in the VENUS critical facility at Mol and open-vessel cold critical experiments performed in the Dodewaard core. The pelleted plutonium rods were fabricated and controlled by Belgonucleaire following the manufacturing procedures developed at the production plant. In one of the assemblies, three vibrated plutonium fuel rods with a lower fuel density were introduced in the three most highly rated positions to reduce the power rating. Those plutonium assemblies experienced peak pellet ratings up to 535 W/cm and were discharged in April 1974 after having reached a mean burnup of approximately 21,000 MWd/MT. In-core instrumentation during operation, visual examinations, and reactivity substitution experiments during reactor shutdown did not indicate any special feature for those assemblies compared to the standard uranium assemblies, thereby demonstrating their interchangeability

  3. Operational experience of decommissioning techniques for research reactors in the United Kingdom

    International Nuclear Information System (INIS)

    England, M.R.; McCool, T.M.

    2002-01-01

    In previous co-ordinated research projects (CRP) conducted by the IAEA no distinction was made between decommissioning activities carried out at nuclear power plants, research reactors or nuclear fuel cycle facilities. As experience was gained and technology advanced it became clear that decommissioning of research reactors had certain specific characteristics which needed a dedicated approach. It was within this context that a CRP on Decommissioning Techniques for Research Reactors was launched and conducted by the IAEA from 1997 to 2001. This paper considers the experience gained from the decommissioning of two research reactors during the course of the CRP namely: (a) the ICI Triga Mk I reactor at Billingham UK which was largely complete by the end of the research project and (b) the Argonaut 100 reactor at the Scottish Universities Research and Reactor centre at East Kilbride in Scotland which is currently is the early stages of dismantling/site operations. It is the intention of this paper with reference to the two case studies outlined above to compare the actual implementation of these works against the original proposals and identify areas that were found to be problematical and/or identify any lessons learnt. (author)

  4. RA reactor operation and maintenance in 1992, Part 1

    International Nuclear Information System (INIS)

    Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Majstorovic, D.; Tanaskovic, M.

    1992-01-01

    During 1992 Ra reactor was not in operation. All the activities were fulfilled according to the previously adopted plan. Basic activities were concerned with revitalisation of the RA reactor and maintenance of reactor components. All the reactor personnel was busy with reconstruction and renewal of the existing reactor systems and building of the new systems, maintenance of the reactor devices. Part of the staff was trained for relevant tasks and maintenance of reactor systems [sr

  5. Surveillance extension experience at WWER-440 type reactors

    International Nuclear Information System (INIS)

    Gillemot, F.; Uri, G.; Oszwald, F.; Trampus, P.

    1993-01-01

    In WWER-440 reactors, the surveillance specimens are located in accelerated irradiation positions. After 5 years, all specimens are withdrawn and the operational changes are not monitored. At Paks NPP a new surveillance program extension has been settled in order to avoid these original program disadvantages and generate further data for plant lifetime management. This paper includes: research performed to prepare the surveillance extension programme, the evaluation method for the surveillance extension, and first results (Charpy and tensile tests). (authors). 6 refs., 12 figs., 3 tabs

  6. Experience with EBR-II [Experimental Breeder Reactor] driver fuel

    International Nuclear Information System (INIS)

    Seidel, B.R.; Porter, D.L.; Walters, L.C.; Hofman, G.L.

    1986-01-01

    The exceptional performance of Experimental Breeder Reactor-II (EBR-II) metallic driver fuel has been demonstrated by the irradiation of a large number of elements under steady-state, transient overpower, and loss-of-flow conditions. High burnup with high reliability has been achieved by a close coupling of element design and materials selection. Quantification of reliability has allowed full utilization of element lifetime. Improved design and duct materials currently under test are expected to increase the burnup from 8 to 14 at.%

  7. Surveillance extension experience at WWER-440 type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gillemot, F; Uri, G [Budapesti Mueszaki Egyetem, Budapest (Hungary); Oszwald, F; Trampus, P

    1994-12-31

    In WWER-440 reactors, the surveillance specimens are located in accelerated irradiation positions. After 5 years, all specimens are withdrawn and the operational changes are not monitored. At Paks NPP a new surveillance program extension has been settled in order to avoid these original program disadvantages and generate further data for plant lifetime management. This paper includes: research performed to prepare the surveillance extension programme, the evaluation method for the surveillance extension, and first results (Charpy and tensile tests). (authors). 6 refs., 12 figs., 3 tabs.

  8. Overview of the FUTURIX-FTA Irradiation Experiment in the Phénix Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Heather J.M. Chichester; Steve L. Hayes; Kenneth J. McClellan; Jean-Luc Paul; Marc Masson; Stewart L. Voit; Fabienne Delage

    2015-09-01

    The Advanced Fuels Campaign utilizes the Advanced Test Reactor (ATR) for most of its irradiation testing. Cadmium-shrouded baskets are used in ATR to modify the neutron spectrum to simulate a fast reactor environment for the fuel. FUTURIX-FTA is an irradiation experiment conducted in the Phenix fast reactor in France. Results from FUTURIX-FTA and irradiation tests in ATR using identical fuel compositions will be compared to identify and evaluate any differences in fuel behavior due to differences in the irradiation source.

  9. Measurement of the thermal flux distribution in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Tangari, C.M.; Moreira, J.M.L.; Jerez, R.

    1986-01-01

    The knowledge of the neutron flux distribution in research reactors is important because it gives the power distribution over the core, and it provides better conditions to perform experiments and sample irradiations. The measured neutron flux distribution can also be of interest as a means of comparison for the calculational methods of reactor analysis currently in use at this institute. The thermal neutron flux distribution of the IEA-R1 reactor has been measured with the miniature chamber WL-23292. For carrying out the measurements, it was buit a guide system that permit the insertion of the mini-chamber i between the fuel of the fuel elements. It can be introduced in two diferent positions a fuel element and in each it spans 26 axial positions. With this guide system the thermal neutron flux distribution of the IEA-R1 nuclear reactor can be obtained in a fast and efficient manner. The element measured flux distribution shows clearly the effects of control rods and reflectors in the IEA-R1 reactor. The difficulties encountered during the measurements are mentioned with detail as well as the procedures adopteed to overcome them. (Author) [pt

  10. Heterogeneous computation tests of both substitution and reactivity worth experiments in the RB-3 reactor

    International Nuclear Information System (INIS)

    Broccoli, U.; Cambi, G.; Vanossi, A.; Zapellini, G.

    1977-01-01

    This report presents the results of several experiments carried out in the D 2 O-moderated RB-3 reactors at the CNEN's Laboratory of Montecuccolino, Bologna. The experiments referred to are either fuel-element substitution experiments or interstitial absorber experiments and were performed during the period 1972-1974. The results of measurements are compared with those obtained by means of computational procedure based on some ''cell'' codes coupled with heterogeneous codes. (authors)

  11. Complete Sensitivity/Uncertainty Analysis of LR-0 Reactor Experiments with MSRE FLiBe Salt and Perform Comparison with Molten Salt Cooled and Molten Salt Fueled Reactor Models

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mueller, Don [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-12-01

    In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in 7Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.

  12. Natural circulation in a VVER reactor geometry: Experiments with the PACTEL facility and Cathare simulations

    Energy Technology Data Exchange (ETDEWEB)

    Raussi, P.; Kainulainen, S. [Lappeenranta Univ. of Technology, Lappeenranta (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)

    1995-09-01

    There are some 40 reactors based on the VVER design in use. Database available for computer code assessment for VVER reactors is rather limited. Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after VVER pressurized water reactors. Flow behaviour over a range of coolant inventories was studied with a small-break experiment. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs. The cause was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. At low primary inventories, core cooling was achieved through the boiler-condenser mode. The experiment was simulated using French thermalhydraulic system code CATHARE.

  13. Benefits of reactor physics experiments for the HTGR industrial development - an attempt to a quantitative approach

    Energy Technology Data Exchange (ETDEWEB)

    Cuniberti, R; Graziani, G; Massino, L; Rinaldini, C; Zanantoni, C

    1972-10-15

    The available results of reactor physics experiments on HTGRs and their accuracies are briefiy reviewed. The physical quantities of interest are grouped into three categories: basic nuclear data, lattice parameters and integral design data. The last two are considered and their possible improvements in accuracy by means of experimental measurements are assessed. The cost penalty on fuel cycle and capital cost due to each physical quantity is then considered, and consequently the benefits of reactor physics experiments are evaluated for a number of hypotheses concerning the foreseeable HTGR development and the delay in taking practical advantage of experimental results. It is concluded that, at the present state of knowledge of basic nuclear data and with the available calculation methods, the economic incentive to new reactor physics experiments is small, and a previous careful analysis is recommended to those intending to perform such experiments.

  14. Thermal and hydraulic characteristics of the JEN-1 Reactor; Caracteristicas hidraulicas y termicas del Reactor JEN-1

    Energy Technology Data Exchange (ETDEWEB)

    Otra Otra, F; Leira Rey, G

    1971-07-01

    In this report an analysis is made of the thermal and hydraulic performances of the JEN-1 reactor operating steadily at 3 Mw of thermal power. The analysis is made separately for the core, main heat exchanger and cooling tower. A portion of the report is devoted to predict the performances of these three main components when and if the reactor was going to operate at a power higher than the maximum 3 Mw attainable today. Finally an study is made of the unsteady operation of the reactor, focusing the attention towards the pumping characteristics and the temperatures obtained in the fuel elements. Reference is made to several digital calculation programmes that nave been developed for such purpose. (Author) 21 refs.

  15. Steam-generator tube performance: world experience with water-cooled nuclear power reactors during 1978

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1980-01-01

    The performance of steam-generator tubes in water-cooled nuclear power reactors during 1978 is reviewed. Tube failures occurred at 31 of the 86 reactors surveyed. The causes of these failures and the procedures designed to deal with them are described. The number of tubes plugged has decreased dramatically in 1978 compared to the previous year. This is attributed to the diligent application of techniques developed through in-plant experience and research and development programs over the past several years

  16. Experience of developing the imitators of the fuel element for the WWER reactors

    International Nuclear Information System (INIS)

    Balashov, S.M.; Boltenko, Eh.A.; Vinogradov, V.A.

    1998-01-01

    Peculiarities of designs of fuel elements imitators for the WWER-type reactors of nominal capacity and with single-ended current feed positioning are considered. The data on the filler heat conductivity and the results of tests and application of the fuel elements imitators at various testing facilities are presented. The possibility of equipping one of the non operating WWER reactors with the fuel element imitators for conduct of large-scale experiment is indicated

  17. Ohmically heated toroidal experiment (OHTE) mobile ignition test reactor facility concept study

    International Nuclear Information System (INIS)

    Masson, L.S.; Watts, K.D.; Piscitella, R.R.; Sekot, J.P.; Drexler, R.L.

    1983-02-01

    This report presents the results of a study to evaluate the use of an existing nuclear test complex at the Idaho National Engineering Laboratory (INEL) for the assembly, testing, and remote maintenance of the ohmically heated toroidal experiment (OHTE) compact reactor. The portable reactor concept is described and its application to OHTE testing and maintenance requirements is developed. Pertinent INEL facilities are described and several test system configurations that apply to these facilities are developed and evaluated

  18. Further analysis of the zero-energy experiment on the Dragon reactor

    International Nuclear Information System (INIS)

    Woloch, F.; Neuberger, W.

    1978-01-01

    The analysis of the Zero-Energy Experiments performed on the Dragon reactor, a high-temperature reactor of the Organization for Economic Cooperation and Development, has been continued. The first analysis established the main route of calculations within the WIMS-E scheme and was reported elsewhere. This Note presents further calculations showing the merits of a refinement in the number of neutron energy groups, of the use of different condensation spectra, and of transport calculations

  19. An evaluation of the results of the HTR fuel programme conducted in the Dragon reactor experiment

    International Nuclear Information System (INIS)

    Shepherd, L.R.

    1982-01-01

    The Dragon Reactor Experiment was used over a period of ten years to investigate the behaviour of HTR fuel elements under realistic service conditions. The purpose of the work was to develop fuel capable of meeting the requirements of commercial power reactors. The studies divided into areas concerned with the mechanical behaviour of the graphite core structure under fast neutron irradiation and the ability of the coated particle fuel to retain fissile products over commercially viable life-cycles. (author)

  20. 14th Biennial conference on reactor operating experience plant operations: The human element

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    Separate abstracts were prepared for the papers presented in the following areas of interest: enhancing operator performance; structured approaches to maintenance standards and reliability-centered maintenance; human issues in plant operations and management; test, research, and training reactor utilization; methods and applications of root-cause analysis; emergency operating procedure enhancement programs; test, research, and training reactor upgrades; valve maintenance and diagnostics; recent operating experiences; and current maintenance issues

  1. IRPhE/RRR-SEG, Reactor Physics Experiments from Fast-Thermal Coupled Facility

    International Nuclear Information System (INIS)

    Weiss, Frank-Peter; Dietze, Klaus; Jacqmin, Robert; Ishikawa, Makoto

    2003-01-01

    1 - Description: The RRR-SEG-experiments have been performed to check neutron data of the most important reactor materials, especially of fission product nuclides, fuel isotopes and structural materials. The measured central reactivity worths (CRW) of small samples were compared with calculated values. These C/E-ratios have been used then for data corrections or in adjustment procedures. The reactor RRG-SEG (at RC Rossendorf / Germany) was a fast-thermal coupled facility of zero power. The annular thermal drivers were filled by fuel assemblies and moderated by water. The inner insertion lattices were loaded with pellets of fuel and other materials producing the fast neutron flux. The characteristics of the neutron and adjoint spectra were obtained by special arrangements of these pellets in unit cells. In this way, a hard or soft neutron spectrum or a special energy behavior of the adjoint function could be reached. The samples were moved by means of tubes to the central position (pile-oscillation technique). The original information about the facility and measurements is compiled in Note Technique SPRC/LEPh/93-230 (SEG) The SEG experiments are considered 'clean' integral experiments, simple and clear in geometry and well suited for calculation. In all SEG configurations only a few materials were used, most of these were standards. Due to the designed adjoint function (energy-independent or monotonously rising), the capture or scattering effect was dominant, convenient to check separately capture or scattering data. At first, analyses of the experiments have been performed in Rossendorf. Newer analyses were done later in Cadarache / CEA France using the European scheme for reactor calculation JEF-2.2 / ECCO / ERANOS (see Note Techniques and JEF/DOC-746). Furthermore, re-analyses were performed in O-arai / JNC Japan with the JNC standard route JENDL-3.2 / SLAROM / CITATION / PERKY. Results obtained with both code systems and different data evaluations (JEF-2.2 and

  2. Deuterium-tritium experiments on the Tokamak Fusion Test reactor

    International Nuclear Information System (INIS)

    Hosea, J.; Adler, J.H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.

    1994-09-01

    The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to ∼9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the PFUS ∼6 MW level. Instability in the TAE mode frequency range has been observed at PFUS > 7 MW and its effect on performance in under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored

  3. Experience on the demonstration of safety for older reactors

    International Nuclear Information System (INIS)

    Facer, R.

    2001-01-01

    The UK's oldest reactors are still operating. Built during the 1950's and commissioned between 1956 and 1960, eight reactors continue to provide electricity and process steam. It is still economically justified to keep them running. In addition to the economic considerations it is also necessary to justify that they can still continue to operate safely. This paper provides a brief review of how the Operator of these stations has justified the safety of operation to date and how they expect to continue to justify their operation for several more years. It is appropriate to consider why the Operator wishes to keep the plant operating. Among the most important reasons are that: The plant is built and paid for, Running costs are relatively low process steam is available for the adjacent sites It is a commercially viable electricity producer It is a reliable electricity source The operators have developed programmes for safety review of the plant and introduced a Continuing Operation Programme which had two main requirements which were, the demonstration of continuing acceptable safety the ensurance of commercial viability. (author)

  4. Reactor Coolant Pump Motor Maintenance Experience in Krsko NPP

    International Nuclear Information System (INIS)

    Vukovic, J.; Besirevic, A.; Boljat, Z.

    2016-01-01

    After thirty years of service as well as maintenance in Krsko NPP both original Reactor Coolant Pump (RCP) motors are remanufactured by original vendor Westinghouse and a new one was purchased. Design function of the RCP motor is to drive Reactor Coolant Pump and for coast-down feature during Design Basis Accident. This paper will give a view on maintenance issues of RCP motor during the thirty years of service and maintenance in Krsko NPP to be kept functionally operational. During the processes of remanufacturing inspection and disassembly it was made possible to get a deeper perspective in the motor condition and the wear or fatigue of the motor parts. Parameters like bearing & winding temperature, absolute and relative vibration greatly affect motor operation if not kept inside design margins. Rotational speed causes heat generation at the bearings which is then associated with oil temperatures and as a consequence bearing temperatures. That is why the most critical parts of the motor are the components of upper and lower bearing assembly. The condition of motor stator and rotor assembly technical characteristics shall be explained with respect to influence of demanding environmental conditions that the motor is exposed. Assessment shall be made how does the wear of critical RCP motor parts can influence reliable performance of the motor if not maintained in proper way. Information on upgrades that were done on RCP motor shall be shared: Oil Spillage Protection System (OSPS), Stator upgrades, Dynamic Port, etc. (author).

  5. Reactor safety instrumentation of Paks NPP (experience and perspective)

    International Nuclear Information System (INIS)

    Elo, S.; Hamar, K.

    1993-01-01

    The majority of the existing control and protection systems in nuclear power plants use old analog technology and design philosophy. Maintenance and the procurement of spare parts is becoming increasingly difficult. In general there is an age degradation concern. Aging degradation in nuclear power plants must be effectively managed to avoid a loss of vital safety function, shutdown of the station, a reduced power generation, or any failure leading to expensive repair. Even with the best efforts in developing reliable and long life instrumentation and control systems for nuclear power plants it is expected that these systems for most plants will require replacements during the life of the plants. The instrumentation and control system of the nuclear power plants designed during the 70's and constructed in the 80's went out-of-date since nuclear safety is not a static concept and the digital computer technology has undergone rapid improvements during the 70's and 80's. Simultaneously the operation and the maintenance of the I ampersand C system of those plants described above becomes more and more difficult and expensive. In this context the pure quality of the former Soviet designed process instrumentation system increases the needs of upgrading this system. The author reviews the main design characteristics of the reactor safety instrumentation of the Paks NPP. Further he attempts to convey the perspective on upgrading the reactor safety instrumentation as seen by the HAEC and its Nuclear Safety Inspectorate

  6. Current activities at the FiR 1 TRIGA reactor

    International Nuclear Information System (INIS)

    Salmenhaara, Seppo

    2002-01-01

    The FiR 1 -reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The epithermal neutrons needed for the irradiation of brain tumor patients are produced from the fast fission neutrons by a moderator block consisting of Al+AlF 3 (FLUENTAL), which showed to be the optimum material for this purpose. Twenty-one patients have been treated since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization. The treatment organization has a close connection to the Helsinki University Central Hospital. The BNCT work dominates the current utilization of the reactor: three days per week for BNCT purposes and only two days per week for other purposes such as the neutron activation analysis and isotope production. In the near future the back end solutions of the spent fuel management will have a very important role in our activities. The Finnish Parliament ratified in May 2001 the Decision in Principle on the final disposal facility for spent fuel in Olkiluoto, on the western coast of Finland. There is a special condition in our operating license. We have now about two years' time to achieve a binding agreement between VTT and the Nuclear Power Plant Companies about the possibility to use the final disposal facility of the Nuclear Power Plants for our spent fuel. If this will not happen, we have to make the agreement with the USDOE with the well-known time limits. At the moment it seems to be reasonable to prepare for both spent fuel management possibilities: the domestic final disposal and the return to the USA offered by USDOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will determine, which of the spent fuel policies will be obeyed. In a couple of years' time it will be seen, if the funding of the reactor and the incomes from the BNC treatments will cover

  7. Evaluation of power behavior during startup and shutdown procedures of the IPR-R1 Triga Reactor

    International Nuclear Information System (INIS)

    Zangirolami, Dante M.; Mesquita, Amir Z.; Ferreira, Andrea V.

    2009-01-01

    The IPR-R1 nuclear reactor of Centro de Desenvolvimento da Tecnologia Nuclear - CDTN/CNEN is a TRIGA Mark I pool type reactor cooled by natural circulation of light water. In the IPR-R1, the power is measured by four nuclear channels, neutron-sensitive chambers, which are mounted around the reactor core: the Startup Channel for power indication during reactor startup; the Logarithmic Wide Range Power Monitoring Channel; the Linear Multi-Range Power Monitoring Channel and the Percent Power Safety Channel. A data acquisition system automatically does the monitoring and storage of all the reactor operational parameters including the reactor power. The startup procedure is manual and the time to reach the desired reactor power level is different on each irradiation which may introduces differences in induced activity of samples irradiated in different irradiations. In this work, the power evolution during startup and shutdown periods of IPR-R1 operation was evaluated and the mean values of reactor energy production in these operational phases were obtained. The analyses were performed on basis of the Linear Multi-Range Channel data. The results show that the sum of startup and shutdown periods corresponds to 1% of released energy for irradiations during 1h at 100kW. This value may be useful to correct experimental data in neutron activation experiments. (author)

  8. Neutron flux measurement and thermal power calibration of the IAN-R1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sarta Fuentes, Jose A.; Castiblanco Bohorquez, Luis A

    2008-10-29

    The IAN-R1 TRIGA reactor in Colombia was initially fueled with MTR-HEU enriched to 93% U-235, operated since 1965 at 10 kW, and was upgraded to 30 kW in 1980. General Atomics achieved in 1997 the conversion of HEU fuel to LEU fuel TRIGA type, and upgraded the reactor power to 100 kW. Since the IAN-R1 TRIGA reactor was in an extended shutdown during seven years, it was necessary to repeat some results of the commissioning test conducted in 1997. The thermal power calibration was carried out using the calorimetric method. The reactor was operated approximately at 20 kW during 3.5 hours, with manual power corrections since the automatic control system failed and with the forced refrigeration off. During the calorimetric experiment, the pool temperature was measured with a RTD which is installed near to the core. The dates were collected in intervals of 30 minutes. For establishing thermal power reactor, the water temperature versus the running were registered. For a calculated tank volume of 16 m{sup 3}, the tank constant calculated for the IAN-R1 TRIGA reactor is 0.0539 C/kW-hr. The reactor power determined was 19 kW. The core configuration is a rectangular grid plate that holds a combination of 4-rod and 3-rod clusters. The core contains 50 fuel rods with LEU fuel TRIGA (UZr H1.6) type enriched to 19.7%. The radial reflector consists of twenty graphite elements six of which are used for isotope production. The top an bottom reflectors are the cylindrical graphite end reflectors which are installed above and below of the active fuel section in each fuel rod. The spatial dependence of thermal neutron flux was measured axially in the 3-rod clusters 4C, 3D, 5E and in the 4F graphite element. The spatial distribution of the thermal neutron was determined using a self-powered detector and the absolute value of thermal neutron flux was determined by a gold activation detector. The (n, b- ) reaction is applied to determine the relative spatial distribution of thermal

  9. Development and Operation of Experiment Course using Research Reactor and Associated Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Shin, B. C.; Hwang, I. A.; Won, J. Y.; Ju, Y. C.; Nam, J. S.; Seo, K. W.; Kim, H. N.

    2013-05-15

    The purpose of present research is to offer a specialized educational opportunity by developing specific curriculum for potential users, mainly university students majoring in related with nuclear engineering and radiation field, on site at KAERI, exploiting the diverse offering of HANARO and ancillary facilities. The specific items of this research accomplished are: First, Development of various curricula for specific research using HANARO and continuous operation of the developed curricula to provided university students with opportunities to use HANARO. Second, Continuous operation of research reactor related experimental training programs for university students in nuclear field to make contribution to cultivating specialists. Third, through the site experimental training for new coming nuclear engineering students, support future potential users to the nuclear research fields, as well as enlarge or broaden the base. Finally, it is hoped that these experiments broadens public awareness and acceptance of the present and potential future contribution of the reactor technology, there by bring positive impacts to policy making. As a whole, 108 students offered and 88 students from 6 universities have completed the course of the programs developed by this project. Also, 1 textbook and 1 teaching aid, a questionnaire have been developed to support the program.

  10. HAV-1-A multipurpose multimonitor for reactor neutron flux characterization

    International Nuclear Information System (INIS)

    Diaz Rizo, O.; Alvarez, I.; Herrera, E.; Lima, L.; Tores, J.; Lopez, M.C.; Ixquiac, M.

    1996-01-01

    A simple method non-solid multi monitor HAV-1 for the systematic evaluation of reactor neutron flux parameters for K o neutron activation analysis is presented. Solution of Au, Zr, Co, Zn, Sn, U and Th (deposited in filter paper) are used to study the parameters alpha and f. Dissolved Lu is used to neutron temperature (Tn) determination, according to the Wescott's formalism. A multipurpose multi monitor HAV-1 preparation, certification and evaluations presented

  11. HAV-1-A multipurpose multimonitor for reactor neutron flux characterization

    Energy Technology Data Exchange (ETDEWEB)

    Diaz Rizo, O; Alvarez, I; Herrera, E; Lima, L; Tores, J [Secretaria Ejecutiva para Asuntos Nucleares, Holguin (Cuba). Delegacion Territorial; Manso, M V [Centro de Isotopos, La Habana (Cuba); Lopez, M C [Instituto Nacional de Investigaciones Nucleares, Mexico City (Mexico); Ixquiac, M [Universidad de San Carlos de Guatemala, Guatemala City (Guatemala)

    1997-12-31

    A simple method non-solid multi monitor HAV-1 for the systematic evaluation of reactor neutron flux parameters for K{sub o} neutron activation analysis is presented. Solution of Au, Zr, Co, Zn, Sn, U and Th (deposited in filter paper) are used to study the parameters alpha and f. Dissolved Lu is used to neutron temperature (Tn) determination, according to the Wescott`s formalism. A multipurpose multi monitor HAV-1 preparation, certification and evaluations presented.

  12. Department of Reactor Technology annual progress report 1 January - 31 December 1978

    International Nuclear Information System (INIS)

    1979-04-01

    The activities of the department of reactor technology at Risoe during 1978 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  13. Verification of the linearity of the IPR-R1 TRIGA reactor power channels

    International Nuclear Information System (INIS)

    Souza, Rose Mary Gomes do Prado; Campolina, Daniel de Almeida Magalhaes

    2013-01-01

    The aim of this paper is to verify the linearity of the three power channels of the IPR-R1 TRIGA reactor. Located at Nuclear Technology Development Center-CDTN in Belo Horizonte, the IPR-R1 reactor is a typical 100 kW Mark I light-water reactor cooled by natural convection. When the experiments were performed, the reactor core had 59 fuel elements, containing 8% by weight of uranium enriched to 20% in 235 U. The core has cylindrical configuration with an annular graphite reflector. The responses of the detectors of the Linear, Log N and Percent Power channels were compared with the responses of detectors which only depend on the overall neutron flux within the reactor. Gold and cobalt foils were activated at low and high powers, respectively, and the specific count results were compared with measurements performed, simultaneously, with a fission chamber, and with the power registered by the three channels. The results show that the Linear channel responds linearly up to 100 kW, and the Log N channel responses are linear at low powers. In the range of high power, the Log N and the Percent Power channels exhibit linearity only from 10 kW to 50 kW. (author)

  14. Utility experience in reactor noise analysis in German LWRs

    International Nuclear Information System (INIS)

    Schuette, A.; Weingarten, J.; Wach, D.

    1988-01-01

    Reactor noise analysis has a long history in Fed. Rep. of Germany. Since the early 60ies it was directed to early fault, malfunction and anomaly detection in nuclear power plants. When successful diagnoses showed the high potential and benefits including economic and safety aspects, also the utilities began to have a vivid concern in application of the developed methods and systems. The know-how transfer from specialized teams to the onsite personnel as well as the development of more ''intelligent'' online systems are important present and future tasks for further improvement of their efficiency. The diagnostic group of GRS is active in both fields and has close cooperation with several utilities to achieve these objectives. Modern signal processing equipment and access to a comprehensive knowledge base stored in data banks are provided by the data analysis center of GRS. The available capabilities are described and examples of analysis results within the cooperation utilities/GRS are given. (author)

  15. Vented fuel experiment for gas-cooled fast reactor application

    International Nuclear Information System (INIS)

    Longest, A.W.; Gat, U.; Conlin, J.A.; Campana, R.J.

    1975-01-01

    A pressure-equalized and vented fuel rod is being irradiated in an instrumented capsule designated GB-10 to approximately 100 MWd/kg-heavy metal. The fuel is a sol-gel derived 88 atom-percent uranium (approximately 9 percent 235 U) 12 atom-percent plutonium oxide, and the cladding is 20 percent cold-worked 316 stainless steel. The capsule is being irradiated in the Oak Ridge Research Reactor (ORR) and has exceeded a burnup of 70 MWd/kg. The fuel has been operated at linear power rates of 39 and 44 kW/ m, and peak outer cladding temperature of 565 0 and 630 0 C respectively. A similar fuel rod in a previous capsule (GB-9) was subjected to 48 kW/m (685 0 C). 4 references. (auth)

  16. Reactor instrumentation experience at IPEN-CNEN/SP

    International Nuclear Information System (INIS)

    Soares, A.J.

    1997-01-01

    The Instituto de Pesquisas Energeticas e Nucleares (IPEN) in Sao Paulo started in 1977 development of nuclear instrumentation. In 1982 two groups were established, one responsible for maintenance of electronic equipment and the other, the Nuclear Instrumentation Section (NIS), for new projects, such as a light water moderated critical facility (IPEN/MB-0). The NIS had the responsibility to execute all necessary activities to develop the ''nonconventional'' instrumentation systems: control, protection and radiation monitoring for the IPEN/MB-01 facility. After the commissioning of the critical facility the NIS group started the development of instrumentation systems for a power reactor. As a first step an overall review of the quality assurance programme was implemented. In parallel the development of self powered detectors was carried out. (author). 7 figs

  17. Sandia reactor kinetics codes: SAK and PK1D

    International Nuclear Information System (INIS)

    Pickard, P.S.; Odom, J.P.

    1978-01-01

    The Sandia Kinetics code (SAK) is a one-dimensional coupled thermal-neutronics transient analysis code for use in simulation of reactor transients. The time-dependent cross section routines allow arbitrary time-dependent changes in material properties. The one-dimensional heat transfer routines are for cylindrical geometry and allow arbitrary mesh structure, temperature-dependent thermal properties, radiation treatment, and coolant flow and heat-transfer properties at the surface of a fuel element. The Point Kinetics 1 Dimensional Heat Transfer Code (PK1D) solves the point kinetics equations and has essentially the same heat-transfer treatment as SAK. PK1D can address extended reactor transients with minimal computer execution time

  18. Reactivity feedback coefficients Pakistan research reactor-1 using PRIDE code

    Energy Technology Data Exchange (ETDEWEB)

    Mansoor, Ali; Ahmed, Siraj-ul-Islam; Khan, Rustam [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Inam-ul-Haq [Comsats Institute of Information Technology, Islamabad (Pakistan). Dept. of Physics

    2017-05-15

    Results of the analyses performed for fuel, moderator and void's temperature feedback reactivity coefficients for the first high power core configuration of Pakistan Research Reactor - 1 (PARR-1) are summarized. For this purpose, a validated three dimensional model of PARR-1 core was developed and confirmed against the reference results for reactivity calculations. The ''Program for Reactor In-Core Analysis using Diffusion Equation'' (PRIDE) code was used for development of global (3-dimensional) model in conjunction with WIMSD4 for lattice cell modeling. Values for isothermal fuel, moderator and void's temperature feedback reactivity coefficients have been calculated. Additionally, flux profiles for the five energy groups were also generated.

  19. Change of neutron flow sensors effectiveness in the course of reactor experiments

    International Nuclear Information System (INIS)

    Kurpesheva, A.M.; Kotov, V.M.; Zhotabaev, Zh.R.

    2007-01-01

    Full text: IGR reactor is a reactor of thermal capacity type. During the operation, uranium-graphite core can be heated up to 1500 deg. C and reactivity can be changed considerably. Core dimensions are comparatively small. Amount of control rods, providing required reactivity, is not big as well. Increasing of core temperature leads to the rise of neutrons path length in its basic material - graphite. Change of temperature is not even. All this causes the non-conservation of neutron flows ratio in irradiated sample and in the place of reactor power sensors installation. Deviations in this ratio were registered during the number of reactor experiments. Empiric corrections can be introduced in order to decrease influence of change of neutron flow effectiveness upon provision of required parameters of investigated matters load. However, dependence of these corrections upon many factors can lead to the increasing of instability of process control. Previous experiment-calculated experiments showed inequality of neutron field in the place of sensors location (up to tens of percent), low effectiveness of experimental works, carried out without access to the individual reactor laying elements. Imperfection during the experiment was an idea of possibility to connect distribution of out of reactor neutron flow and control rods position. Subsequent analysis showed that for the development of representative phenomenon model it is necessary to take into account reactor operation dynamic subject to unevenness of heating of individual laying parts. Elemental calculations showed that temperature laying effects in the change of neutron outer field are great. Algorithm of calculations for the change of outer filed and field of investigated fabrication includes calculation of neutron-physic reactor characteristics interlacing with calculations of thermal-physic reactor characteristics, providing correlation of temperature fields for neutron-physic calculations. In the course of such

  20. FLOW TESTING AND ANALYSIS OF THE FSP-1 EXPERIMENT

    Energy Technology Data Exchange (ETDEWEB)

    Hawkes, Grant L.; Jones, Warren F.; Marcum, Wade; Weiss, Aaron; Howard, Trevor

    2017-06-01

    The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A flow versus delta P curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported herein. Fuel plate deformation tests were also performed and reported.

  1. Reactor core conversion studies of Ghana: Research Reactor-1 and proposal for addition of safety rod

    International Nuclear Information System (INIS)

    Odoi, H.C.

    2014-06-01

    The inclusion of an additional safety rod in conjunction with a core conversion study of Ghana Research Reactor-1 (GHARR-1) was carried out using neutronics, thermal hydraulics and burnup codes. The study is based on a recommendation by Integrated Safety Assessment for Research Reactors (INSARP) mission to incorporate a safety rod to the reactor safety system as well as the need to replace the reactor fuel with LEU. Conversion from one fuel type to another requires a complete re-evaluation of the safety analysis. Changes to the reactivity worth, shutdown margin, power density and material properties must be taken into account, and appropriate modifications made. Neutronics analysis including burnup was studied followed by thermal hydraulics analyses which comprise steady state and transients. Four computer codes were used for the analysis; MCNP, REBUS, PLTEP and PARET. The neutronics analysis revealed that the LEU core must be operated at 34 Kw in order to attain the flux of 1.0E12 n/cm 2 .s as the nominal flux of the HEU core. The auxiliary safety rod placed at a modified irradiation site gives a better worth than the cadmium capsules. For core excess reactivity of 4 mk, 348 fuel pins would be appropriate for the GHARR-1 LEU core. Results indicate that flux level of 1.0E12 n/cm 2 .s in the inner irradiation channel will not be compromised, if the power of the LEU core is increased to 34 kW. The GHARR-1 core using LEU-U0 2 -12.5% fuel can be operated for 23 shim cycles, with cycles length 2.5 years, for over 57 years at the 17 kW power level. All 23 LEU cycles meet the ∼ 4.0 mk excess reactivity required at the beginning of cycle . For comparison, the MNSR HEU reference core can also be operated for 23 shim cycles, but with a cycle length of 2.0 years for just over 46 years at 15.0kW power level. It is observed that the GHARR-1 core with LEU UO 2 fuel enriched to 12.5% and a power level of 34 kW can be operated ∼25% longer than the current HEU core operated at

  2. Operating experience and maintenance at the TRIGA Mark II LENA reactor

    International Nuclear Information System (INIS)

    Cingoli, F.; Meloni, S.; Alloni, L.

    1986-01-01

    A summary of reactor operation and maintenance in the time period 1982-1986 is presented and discussed. Some problems occurred from instrumentated aluminum cladded elements. Both of them presented damage in the cable tubes and one element showed a protuberance in the cladding. They were replaced with stainless - steel cladded ones. Both elements were sealed up in stainless - steel tubes and put away in wells, 3 meters deep, in the reactor room floor. Some minor problems, correlated to the quite aid instrumentation of the console, are reported. The reactor activity in the last four years was conditioned by the developing of the n - n-bar oscillation NADIR experiment. The thermal column was dismantled and rebuilt in consideration of the Nadir experiment necessities and this job is described in detail. The building containing, the target and the void pipe, presented in 1982 Conference, are now completely operating and the experiment is running. (author)

  3. China: EDF's feedback experience of reactor operating is essential to win international markets

    International Nuclear Information System (INIS)

    Maillart, H.

    2016-01-01

    The main assets of EDF on the Chinese nuclear power market is first, its very important feedback experience of reactor operations (EDF cumulates one year of reactor operations every week due to its fleet of 58 reactors), secondly the cooperation with China allowed China to enter nuclear energy in 1983 with the construction of the Daya Bay plant and now to develop its own technology: the CPR-1000 reactor. China is the world leader in terms of nuclear market dynamism with 30 reactors in operation, 24 reactors being built and 40 others planned. A new stage in the Franco-China cooperation would be to share relevant good practices in the managing of both French and Chinese fleets of reactors. EDF has upgraded its commercial international offer, it now proposes to cover all the stages of the nuclear power plant from site selection to plant deconstruction via construction, operation, maintenance and waste management which constitutes a commitment over a 100 year period. (A.C.)

  4. Preparation of a criticality benchmark based on experiments performed at the RA-6 reactor

    International Nuclear Information System (INIS)

    Bazzana, S.; Blaumann, H; Marquez Damian, J.I

    2009-01-01

    The operation and fuel management of a reactor uses neutronic modeling to predict its behavior in operational and accidental conditions. This modeling uses computational tools and nuclear data that must be contrasted against benchmark experiments to ensure its accuracy. These benchmarks have to be simple enough to be possible to model with the desired computer code and have quantified and bound uncertainties. The start-up of the RA-6 reactor, final stage of the conversion and renewal project, allowed us to obtain experimental results with fresh fuel. In this condition the material composition of the fuel elements is precisely known, which contributes to a more precise modeling of the critical condition. These experimental results are useful to evaluate the precision of the models used to design the core, based on U 3 Si 2 and cadmium wires as burnable poisons, for which no data was previously available. The analysis of this information can be used to validate models for the analysis of similar configurations, which is necessary to follow the operational history of the reactor and perform fuel management. The analysis of the results and the generation of the model were done following the methodology established by International Criticality Safety Benchmark Evaluation Project, which gathers and analyzes experimental data for critical systems. The results were very satisfactory resulting on a value for the multiplication factor of the model of 1.0000 ± 0.0044, and a calculated value of 0.9980 ± 0.0001 using MCNP 5 and ENDF/B-VI. The utilization of as-built dimensions and compositions, and the sensitivity analysis allowed us to review the design calculations and analyze their precision, accuracy and error compensation. [es

  5. New Reactor Siting, Licensing and Construction Experience. Proceedings of the 2. CNRA International Workshop on 'New Reactor Siting, Licensing and Construction Experience'

    International Nuclear Information System (INIS)

    2013-01-01

    This report documents the proceedings from the 2. Workshop on New Reactor Siting, Licensing and Construction Experience. A total of 45 specialists from 16 countries and international organisations attended. The meeting was sponsored by the OECD Nuclear Energy Agency Committee on Nuclear Regulatory Activities and hosted by the US Nuclear Regulatory Commission (U.S.NRC). The objectives of the workshop were to provide a forum to exchange information on lessons learned from siting, licensing and constructing new nuclear power plants around the world. Key focus areas included siting practices and regulatory positions that have been enhanced as a result of the Fukushima accident; lessons learned from licensing and design review approaches and challenges, construction experience and recommendations for regulatory oversight; and regulatory cooperation on generic and design specific issues through the MDEP specific working groups. The workshop was structured in 4 technical sessions, each followed by ample time for panel discussions. The first technical session was devoted to regulatory cooperation on generic and design specific issues, MDEP working groups (EPR, AP1000), vendor inspection co-operation, digital I and C, and codes and standards. The second technical session was intended to discuss and share regulatory positions on siting practices and enhancements as a result of lessons learned from Fukushima accident. The third technical session addressed the construction experience and regulatory oversight of new reactor construction activities. And the fourth technical session included presentations on the lessons learned from regulatory licensing reviews of new reactor designs

  6. Water treatment process in the JEN-1 Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Urgel, M; Perez-Bustamante, J A; Batuecas, T

    1965-07-01

    The main characteristics and requirements which must be met with by waters to be used for nuclear reactors were studied paying attention separately both to those used in primary and secondary circuits as well as to the purification systems to be employed in each case. The experiments carried out for the initial pretreatment of water and the ion-exchange de ionization processes including a number of systems consisting of separated and mixed beds loaded with a variety of different commercially available resins are described. (Author) 24 refs.

  7. Water treatment process in the JEN-1 Research Reactors

    International Nuclear Information System (INIS)

    Urgel, M.; Perez-Bustamante, J. A.; Batuecas, T.

    1965-01-01

    The main characteristics and requirements which must be met with by waters to be used for nuclear reactors were studied paying attention separately both to those used in primary and secondary circuits as well as to the purification systems to be employed in each case. The experiments carried out for the initial pretreatment of water and the ion-exchange de ionization processes including a number of systems consisting of separated and mixed beds loaded with a variety of different commercially available resins are described. (Author) 24 refs

  8. Evaluation of performance of select fusion experiments and projected reactors. Final report

    International Nuclear Information System (INIS)

    Miley, G.H.

    1978-10-01

    The performance of NASA Lewis fusion experiments (SUMMA and Bumpy Torus) is compared with other experiments and that necessary for a power reactor. Key parameters cited are gain (fusion power/input power) and the time average fusion power, both of which may be more significant for real fusion reactors than the commonly used Lawson parameter. The NASA devices are over 10 orders of magnitude below the required powerplant values in both gain and time average power. The best experiments elsewhere are also as much as 4 to 5 orders of magnitude low. However, the NASA experiments compare favorably with other alternate approaches that have received less funding than the mainline experiments. The steady-state character and efficiency of plasma heating are strong advantages of the NASA approach. The problem, though, is to move ahead to experiments of sufficient size to advance in gain and average power parameters

  9. Update on the direct n-n scattering experiment at the reactor YAGUAR

    Science.gov (United States)

    Stephenson, S. L.; Crawford, B. E.; Furman, W. I.; Lychagin, E. V.; Muzichka, A. Yu.; Nekhaev, G. V.; Sharapov, E. I.; Shvetsov, V. N.; Strelkov, A. V.; Levakov, B. G.; Lyzhin, A. E.; Chernukhin, Yu. I.; Howell, C. R.; Mitchell, G. E.; Tornow, W.; Showalter-Bucher, R. A.

    2013-10-01

    The first direct measurement of the 1S0 neutron-neutron scattering experiment using the YAGUAR aperiodic reactor at the Russian Federal Nuclear Center - All Russian Research Institute of Technical Physics has preliminary results. Thermal neutrons are scattered from a thermal neutron ``gas'' within the scattering chamber of the reactor and measured via time-of-flight. These initial results show an unexpectedly large thermal neutron background now understood to be from radiation-induced desorption within the scattering chamber. Analysis of the neutron time-of-flight spectra suggests neutron scattering from H2 and possibly H2O molecules. An experimental value for the desorption yield ηγ of 0.02 molecules/gamma agrees with modeled results. Techniques to reduce the effect of the nonthermal desorption will be presented. This work was supported in part by ISTC project No. 2286, Russia Found. Grant 01-02-17181, the US DOE grants Nos. DE-FG02-97-ER41042 and DE-FG02-97-ER41033, and by the US NSF through Award Nos. 0107263 and 0555652.

  10. Operating experience of natural circulation core cooling in boiling water reactors

    International Nuclear Information System (INIS)

    Kullberg, C.; Jones, K.; Heath, C.

    1993-01-01

    General Electric (GE) has proposed an advanced boiling water reactor, the Simplified Boiling Water Reactor (SBWR), which will utilize passive, gravity-driven safety systems for emergency core coolant injection. The SBWR design includes no recirculation loops or recirculation pumps. Therefore the SBWR will operate in a natural circulation (NC) mode at full power conditions. This design poses some concerns relative to stability during startup, shutdown, and at power conditions. As a consequence, the NRC has directed personnel at several national labs to help investigate SBWR stability issues. This paper will focus on some of the preliminary findings made at the INEL. Because of the broad range of stability issues this paper will mainly focus on potential geysering instabilities during startup. The two NC designs examined in detail are the US Humboldt Bay Unit 3 BWR-1 plant and Dodewaard plant in the Netherlands. The objective of this paper will be to review operating experience of these two plants and evaluate their relevance to planned SBWR operational procedures. For completeness, experimental work with early natural circulation GE test facilities will also be briefly discussed

  11. Design, construction and operating experience of demonstration LMFBRs. The application of core and fuel performance experience in British reactors to commercial fast reactor design

    International Nuclear Information System (INIS)

    Bagley, K.Q.

    1978-01-01

    The Prototype Fast Reactor (PFR) sub-assembly design is described with particular emphasis on the choice of factors that are important in determining satisfactory performance. Reasons for the adoption of specific clad and fuel design details are given in their historical context, and irradiation experience - mostly from the Dounreay Fast Reactor (DFR) - in support of the choices is described. The implications of factors that are now better understood than when the PFR fuel was designed, notably neutron-induced void swelling and irradiation creep, are then considered. It is shown that the 'free-standing' core design used in PFR, in which the sub-assembly is unsupported above the level of the lower axial breeder, relies on the availability of low-swelling, preferably irradiation-creep-resistant alloys as sub-assembly structural materials in order to achieve the prescribed burn-up target. The advantages of a 'restrained core', which makes use of irradiation creep to redress the effects of material swelling, are noted briefly, and the application of this concept to the Commercial Demonstration Fast Reactor (CDFR) core design is described. Probable future trends in pin and sub-assembly design are reviewed and the scope of associated irradiation testing programmes defined. Arrangements for monitoring and evaluating fuel performance, both in reactor and post-irradiation, are outlined and the provisions for endorsement of CDFR pin, sub-assembly and core design details in PFR are indicated. (author)

  12. Shielding assessment of the ETRR-1 Reactor Under power upgrading

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, E E [Reactor Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    The assessment of existing shielding of the ETRR-1 reactor in case of power upgrading is presented and discussed. It was carried out using both the present EK-10 type fuel elements and some other types of fuel elements with different enrichments. The shielding requirements for the ETRR-1 when power is upgraded are also discussed. The optimization curves between the upgraded reactor power and the shield thickness are presented. The calculation have been made using the ANISN code with the DLC-75 data library. The results showed that the present shield necessitates an additional layer of steel with thickness of 10.20 and 25 cm. When its power is upgraded to 3, 6 and 10 MWt in order to cutoff all neutron energy groups to be adequately safe under normal operating conditions. 4 figs.

  13. Experiment monitoring system of a new electromagnet drive for nuclear reactor control rod

    International Nuclear Information System (INIS)

    Zhang Jige; Wang Xiaoguang; Wu Yuanqiang; Zhang Zhengming

    2003-01-01

    In order to deal with some unsolved problems in the engineering prototype design of a new electromagnet drive device for nuclear reactor control rod, the property experiment in view of principle prototype is carried out. Actual displacement of nuclear reactor control rod is measured by means of raster ruler and the test data is obtained by means of computer. The computer communicates with PLC using RS232 serial port. The experimental results show that the monitoring system have the properties of high reliability and high precision, and ensures the experiment to accomplish successfully

  14. Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

    International Nuclear Information System (INIS)

    Song, C. H.; Baek, W. P.; Chung, M. K.

    2007-06-01

    The objectives of the project are to study thermal hydraulic characteristics of advanced nuclear reactor system for evaluating key thermal-hydraulic phenomena relevant to new safety concepts. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. The Followings are main research topics: - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation-induced Thermal Mixing in a Pool - Development of Thermal-Hydraulic Models for Two-Phase Flow - Construction of T-H Data Base

  15. Quality assurance measures at the Geesthacht research reactor FRG-1

    International Nuclear Information System (INIS)

    Voss, J.; Krull, W.; Schmidt, K.

    1995-01-01

    The major part of the quality system for the FRG is already in practical use; other parts require extensive preparations and therefore transition periods of different lengths before they can be introduced. GKSS is aware that beside the other upgrading and refurbishment activities, these duality assurance measures will be very important in ensuring the operation of the FRG-1 research reactor over the coming 15 years or more. (orig.)

  16. UPGRADE OF INSTRUMENTATION FOR PURDUE REACTOR PUR-1, PHASE 3

    International Nuclear Information System (INIS)

    Revankar, S. T.

    2004-01-01

    The major objective of this program is to upgrade and replace instruments and equipment that significantly improve the performance, control and operational capability of the Purdue University nuclear reactor (PUR-1). Under this major objective one project on design and installation of interface cards for channel four detector was considered. This report is the final report and gives the efforts and progress achieved on these projects from August 2002 to July 2004

  17. Real time monitoring system of the operation variables of the TRIGA IPR-R1 nuclear research reactor

    International Nuclear Information System (INIS)

    Ricardo, Carla Pereira; Mesquita, Amir Zacarias

    2007-01-01

    During the last two years all the operation parameters of the TRIGA IPR-R1 were monitored and real time indicated bu the data acquisition system developed for the reactor. All the information were stored on a rigid disk, at the collection system computer, leaving the information on the reactor performance and behaviour available for consultation in a chronological order. The data acquisition program has been updated and new reactor operation parameters were included for increasing the investigation and experiments possibilities. The register of reactor operation variables are important for the immediate or subsequent safety analyses for reporting the reactor operations to the external organizations. This data acquisition satisfy the IAEA recommendations. (author)

  18. The first critical experiment with a new type of fuel assemblies IRT-3M on the training reactor VR-I

    International Nuclear Information System (INIS)

    Matejka, Karel; Sklenka, Lubomir

    1997-01-01

    The paper 'The first critical experiment with a new type of fuel assemblies IRT-3M on training reactor VR-1 presents basic information about the replacement of fuel on the reactor VR-1 run on FJFI CVUT in Prague. In spring 1997 the IRT-2M fuel type used till then was replaced by the IRT-3M type. When the fuel was replaced, no change in its enrichment was made, i.e. its level remained as 36% 235 U. The replacement itself was carried out in tight co-operation with the Nuclear Research Institute Rez plc., as related to the operation of the research reactor LVR-15. The fuel replacement on the VR-I reactor is a part of the international program RERTR (Reduced Enrichment for Research and Test Reactors) in which the Czech Republic participates. (author)

  19. Thorium fuels for heavy water reactors. Romanian experience

    International Nuclear Information System (INIS)

    Glodeanu, F.; Mirion, I.; Mehedinteanu, S.; Balan, V.

    1984-01-01

    The renewed interest in thorium fuel cycle due to the increased demand for fissile materials has resulted in speeding up the related research and development activities. For heavy water reactors the thorium cycles, especially SSET, are very promising and many efforts are made to demonstrate their feasibility. In our country, at INPR, the research and development activity has been initiated in the following areas: the conceptual design of thorium bearing fuel elements; fuel modelling; nuclear grade thorium dioxide powder technology; mixed oxide fuel technology. In the design area, the key factors in performance limitation, especially at extended burnup have been accounted and different remedies proposed. An irradiation programme has been settled and will start this year. The modelling activities are focused on mixed oxide behaviour and material data measurements are in progress. In the nuclear grade thorium powder technology area, a good piece of work has been done to develop an integrated technology for monasite processing (thorium being a by-product in lanthanides extraction). As regards the mixed oxide fuel technology, efforts have been made to obtain (ThU)O 2 pellets with good homogeneity and high density at different compositions. Besides the mixing powders route, other non-conventional technologies for refabrication like: microspheres, pellet impregnation and clay extrusion are studied. Experimental fuel rods for irradiation testing have been manufactured. (author)

  20. CANDU fuel - fifteen years of power reactor experience

    International Nuclear Information System (INIS)

    Fanjoy, G.R.; Bain, A.S.

    1977-05-01

    Analyses of performance statistics, supplemented by examinations of fuel from power reactors and experimental loops have yielded: (a) a thorough understanding of the fundamental behaviour of CANDU fuel; (b) data showing that the predicted high utilization of uranium has been achieved; (c) criteria for operation, which have led to the current very low defect rate of 0.03% of all assemblies and to 'CANLUB' fuel, which has a graphite interlayer between the fuel and sheath to reduce defects on power increases; (d) proof that the short length (500 mm), collapsible cladding features of the CANDU bundle are successful and that the fuel can operate at high-power output (current peak outer-element linear power is 58 +- 15% kW/m). As of mid-1976 over 3 x 10 6 individual elements have been built and over 2 x 10 6 elements have been irradiated. Only six defects have been attributed to faulty materials or fabrication, and the use of high-density UO 2 with low-moisture content precluded defects from hydrogen contamination and densification